CN116426756A - Method for efficiently leaching uranium and zirconium from uranium purification waste - Google Patents
Method for efficiently leaching uranium and zirconium from uranium purification waste Download PDFInfo
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- CN116426756A CN116426756A CN202310201966.0A CN202310201966A CN116426756A CN 116426756 A CN116426756 A CN 116426756A CN 202310201966 A CN202310201966 A CN 202310201966A CN 116426756 A CN116426756 A CN 116426756A
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- 229910052770 Uranium Inorganic materials 0.000 title claims abstract description 292
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 title claims abstract description 290
- 238000002386 leaching Methods 0.000 title claims abstract description 233
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 title claims abstract description 154
- 229910052726 zirconium Inorganic materials 0.000 title claims abstract description 153
- 239000002699 waste material Substances 0.000 title claims abstract description 143
- 238000000746 purification Methods 0.000 title claims abstract description 109
- 238000000034 method Methods 0.000 title claims abstract description 73
- HEMHJVSKTPXQMS-UHFFFAOYSA-M Sodium hydroxide Chemical compound [OH-].[Na+] HEMHJVSKTPXQMS-UHFFFAOYSA-M 0.000 claims abstract description 210
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- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 claims abstract description 96
- 239000003513 alkali Substances 0.000 claims abstract description 82
- 239000002893 slag Substances 0.000 claims abstract description 62
- 238000002156 mixing Methods 0.000 claims abstract description 16
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Chemical compound O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims description 43
- 239000008367 deionised water Substances 0.000 claims description 24
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- 238000005119 centrifugation Methods 0.000 description 4
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- 150000002500 ions Chemical class 0.000 description 4
- 239000000463 material Substances 0.000 description 4
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- RKTYLMNFRDHKIL-UHFFFAOYSA-N copper;5,10,15,20-tetraphenylporphyrin-22,24-diide Chemical group [Cu+2].C1=CC(C(=C2C=CC([N-]2)=C(C=2C=CC=CC=2)C=2C=CC(N=2)=C(C=2C=CC=CC=2)C2=CC=C3[N-]2)C=2C=CC=CC=2)=NC1=C3C1=CC=CC=C1 RKTYLMNFRDHKIL-UHFFFAOYSA-N 0.000 description 3
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- 229910018072 Al 2 O 3 Inorganic materials 0.000 description 1
- BHPQYMZQTOCNFJ-UHFFFAOYSA-N Calcium cation Chemical compound [Ca+2] BHPQYMZQTOCNFJ-UHFFFAOYSA-N 0.000 description 1
- ZAMOUSCENKQFHK-UHFFFAOYSA-N Chlorine atom Chemical compound [Cl] ZAMOUSCENKQFHK-UHFFFAOYSA-N 0.000 description 1
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- 241000282376 Panthera tigris Species 0.000 description 1
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- XLYOFNOQVPJJNP-UHFFFAOYSA-M hydroxide Chemical compound [OH-] XLYOFNOQVPJJNP-UHFFFAOYSA-M 0.000 description 1
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- 230000027756 respiratory electron transport chain Effects 0.000 description 1
- 230000000630 rising effect Effects 0.000 description 1
- -1 silicate ion Chemical class 0.000 description 1
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- 239000002910 solid waste Substances 0.000 description 1
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- 238000002076 thermal analysis method Methods 0.000 description 1
- 238000012876 topography Methods 0.000 description 1
- 239000010863 uranium mill tailing Substances 0.000 description 1
- NBWXXYPQEPQUSB-UHFFFAOYSA-N uranium zirconium Chemical compound [Zr].[Zr].[U] NBWXXYPQEPQUSB-UHFFFAOYSA-N 0.000 description 1
- 238000004876 x-ray fluorescence Methods 0.000 description 1
- GFQYVLUOOAAOGM-UHFFFAOYSA-N zirconium(iv) silicate Chemical group [Zr+4].[O-][Si]([O-])([O-])[O-] GFQYVLUOOAAOGM-UHFFFAOYSA-N 0.000 description 1
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- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B7/00—Working up raw materials other than ores, e.g. scrap, to produce non-ferrous metals and compounds thereof; Methods of a general interest or applied to the winning of more than two metals
- C22B7/006—Wet processes
- C22B7/007—Wet processes by acid leaching
-
- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B1/00—Preliminary treatment of ores or scrap
- C22B1/005—Preliminary treatment of scrap
-
- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B34/00—Obtaining refractory metals
- C22B34/10—Obtaining titanium, zirconium or hafnium
- C22B34/14—Obtaining zirconium or hafnium
-
- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B60/00—Obtaining metals of atomic number 87 or higher, i.e. radioactive metals
- C22B60/02—Obtaining thorium, uranium, or other actinides
- C22B60/0204—Obtaining thorium, uranium, or other actinides obtaining uranium
- C22B60/0217—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes
- C22B60/0221—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching
- C22B60/0226—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching using acidic solutions or liquors
- C22B60/0239—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching using acidic solutions or liquors nitric acid containing ion as active agent
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02P—CLIMATE CHANGE MITIGATION TECHNOLOGIES IN THE PRODUCTION OR PROCESSING OF GOODS
- Y02P10/00—Technologies related to metal processing
- Y02P10/20—Recycling
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- General Life Sciences & Earth Sciences (AREA)
- Manufacturing & Machinery (AREA)
- Materials Engineering (AREA)
- Mechanical Engineering (AREA)
- Life Sciences & Earth Sciences (AREA)
- Organic Chemistry (AREA)
- Geology (AREA)
- Environmental & Geological Engineering (AREA)
- Geochemistry & Mineralogy (AREA)
- Processing Of Solid Wastes (AREA)
Abstract
The invention provides a method for efficiently leaching uranium and zirconium from uranium purification waste, which comprises the following steps: mixing uranium purification waste with sodium hydroxide solution, pretreating, and separating to obtain alkali treatment slag; and mixing the alkali treatment slag with a nitric acid solution, and leaching to obtain a leaching solution containing uranium and zirconium. The application adopts a two-step leaching method, firstly, the silicon mineral component in uranium purification waste can be effectively crushed by pretreatment with NaOH solution, and mineral matrix is dissolved to generate a large amount of amorphous Zr (OH) 4 So that zirconium and SiO 2 The interlocking relationship between the two is destroyed to form a loose porous structure, the uranium phase is destroyed by the wrapped structure, the uranium is exposed, and the uranium is realizedHigh-efficiency leaching of elements, and Zr (OH) generated 4 Is attached to the surface of the alkali-treated slag and further reacts with nitric acid solution in the subsequent nitric acid solution leaching process to generate ZrO (NO) 3 ) 2 The target element zirconium is dissolved in nitric acid solution and is leached out efficiently.
Description
Technical Field
The invention belongs to the technical field of resource recovery, and particularly relates to a method for efficiently leaching uranium and zirconium from uranium purification waste.
Background
Uranium and zirconium are widely used as strategic resources in the fields of scientific research, industry, energy sources and the like. The natural uranium is taken as a dual-purpose resource for military and civil, so that the stable supply of nuclear power is concerned, and the effectiveness of national nuclear deterrence is fundamentally ensured. Zirconium with outstanding nuclear properties is an indispensable material for the development of the atomic energy industry because of its small thermal neutron capture cross section. However, as non-renewable resources, traditional single uranium and zirconium resources are becoming scarce. According to the nuclear power development planning of China, the accumulated metal uranium demand from 2016 to 2023 reaches approximately 25 ten thousand tons, and the corresponding zirconium demand is multiplied. This means that the supply shortage of metallic uranium, zirconium will be directly hampered to the development of the nuclear industry. Therefore, recycling the uranium and valuable component zirconium which can be reused from the nuclear industry waste residues becomes a trend of guaranteeing natural uranium and zirconium resources.
For uranium and zirconium resource extraction, many scholars have made a constant search in this respect. Lei et al found that by CaCl 2 The roasting pretreatment changes the mineral phase of uranium tailings, and uses calcium element to replace silicon element, so that the formed Ca=O bond strength is lower than Si=O bond strength, silicate can react with acid more easily, and removal of uranium and other radionuclides during acid leaching is facilitated. As zirconium metal resistant to chemical decomposition and corrosion, research such as Reginaldo Jose Farias da Silva and the like utilizes the intense chemical reaction of NaOH alkali fusion to stimulate and destroy zircon structure, generates fusion with the characteristics of high porosity, high fragility and strong corrosiveness, leaches the fusion by a water leaching method, further removes insoluble matters in the solution by a sulfuric acid leaching method, and finally precipitates and dries to form ZrO 2 . The uranium and zirconium extraction methods are limited to single methodThe recovery of resources, namely the collaborative extraction of uranium and zirconium resources, has not been reported at present. Therefore, how to efficiently extract uranium and zirconium from uranium purification waste at the same time is a problem to be solved in the art.
Disclosure of Invention
The invention aims to provide a method for efficiently leaching uranium and zirconium from uranium purification waste. The method provided by the invention can leach uranium and zirconium at the same time, and has high leaching rate.
In order to achieve the above object, the present invention provides the following technical solutions:
the invention provides a method for efficiently leaching uranium and zirconium from uranium purification waste, which comprises the following steps:
(1) Mixing uranium purification waste with sodium hydroxide solution, pretreating, and separating to obtain alkali treatment slag;
(2) And (3) mixing the alkali treatment slag obtained in the step (1) with a nitric acid solution, and leaching to obtain a leaching solution containing uranium and zirconium.
Preferably, the concentration of the sodium hydroxide solution in the step (1) is 80-200 g/L, and the mass ratio of the volume of the sodium hydroxide solution to the uranium purification waste material is (3-7) mL:1g.
Preferably, the temperature of the pretreatment in the step (1) is room temperature, and the pretreatment time is 0.5-3 h.
Preferably, the pretreatment in the step (1) is performed under stirring conditions, and the stirring speed is 200-400 rpm.
Preferably, the product separated in the step (1) is washed with deionized water.
Preferably, the mass ratio of the volume of deionized water to the separated product is (1-13) mL:1g.
Preferably, in the step (2), the volume ratio of nitric acid to water in the nitric acid solution is (0.4-1.3): 1.
preferably, the mass ratio of the volume of the nitric acid solution to the alkali treatment slag in the step (2) is (3-7) mL:1g.
Preferably, the leaching temperature in the step (2) is 20-100 ℃, and the leaching time is 2-2.5 h.
Preferably, the leaching temperature in the step (2) is 50-80 ℃.
The invention provides a method for efficiently leaching uranium and zirconium from uranium purification waste, which comprises the following steps: mixing uranium purification waste with sodium hydroxide solution, pretreating, and separating to obtain alkali treatment slag; and mixing the alkali treatment slag with a nitric acid solution, and leaching to obtain a leaching solution containing uranium and zirconium. The application adopts a two-step leaching method, firstly, the silicon mineral component in uranium purification waste can be effectively crushed by pretreatment with NaOH solution, and mineral matrix is dissolved to generate a large amount of amorphous Zr (OH) 4 So that zirconium and SiO 2 The interlocking relationship is destroyed, thereby forming a loose porous structure, destroying the wrapped uranium phase, exposing the uranium, providing a path for leaching uranium by the nitric acid solution, thereby realizing the efficient leaching of uranium element and generating Zr (OH) 4 Is attached to the surface of the alkali-treated slag and further reacts with nitric acid solution in the subsequent nitric acid solution leaching process to generate ZrO (NO) 3 ) 2 The precipitate is caused to be dissolved, and the target element zirconium is dissolved in nitric acid solution to be leached out efficiently. Experimental results show that the leaching rate of uranium reaches 99.37% and the leaching rate of zirconium reaches 97.48%.
Drawings
FIG. 1 shows the leaching rate and content of uranium in the leaching solutions prepared in examples 1 to 7;
FIG. 2 shows the leaching rate and content of zirconium in the leaching solutions prepared in examples 1 to 7;
FIG. 3 shows the leaching rate and content of uranium in the leaching solutions prepared in example 1 and examples 8 to 11;
FIG. 4 shows the leaching rate and content of zirconium in the leaching solutions prepared in example 1 and examples 8 to 11;
FIG. 5 shows the leaching rate and content of uranium in the leaching solutions prepared in example 1 and examples 12 to 15;
FIG. 6 shows the leaching rate and content of zirconium in the leaching solutions prepared in example 1 and examples 12 to 15;
FIG. 7 shows the leaching rate and content of uranium in the leaching solutions prepared in example 1 and examples 16 to 19;
FIG. 8 shows the leaching rate and content of zirconium in the leaching solutions prepared in example 1 and examples 16 to 19;
FIG. 9 shows the pH of deionized water after washing in examples 1 and 16-19;
FIG. 10 shows the leaching rates and contents of uranium in the leaching solutions prepared in examples 20 to 23 and comparative example 1;
FIG. 11 shows the leaching rates and contents of zirconium in the leaching solutions prepared in examples 20 to 23 and comparative example 1;
FIG. 12 shows the leaching rates and contents of uranium in the leaching solutions prepared in example 20, examples 24 to 26 and comparative example 2;
FIG. 13 shows the leaching rates and contents of zirconium in the leaching solutions prepared in example 20, examples 24 to 26 and comparative example 2;
fig. 14 shows the leaching rate and content of uranium in the leaching solutions prepared in example 20 and examples 27 to 30;
FIG. 15 shows the leaching rate and content of zirconium in the leaching solutions prepared in example 20 and examples 27 to 30;
FIG. 16 is a carbon summary of uranium purified waste material, alkali treated slag, and nitric acid solution leached uranium purified waste material of example 20;
FIG. 17 is a uranium spectrum of uranium purified waste material, alkali treated slag, and nitric acid solution leached uranium purified waste material of example 20;
FIG. 18 is a zirconium spectrogram of the uranium purified waste material, alkali treated slag and nitric acid solution leached uranium purified waste material of example 20;
FIG. 19 is an XRD diffraction pattern of uranium purified waste material, alkali treated slag, and nitric acid solution leached uranium purified waste material of example 20;
FIG. 20 is a TG-DEC curve of alkali-treated slag of example 20;
FIG. 21 is an XRD diffraction pattern of the alkali treated slag of example 20 after roasting at 300 ℃, 600 ℃ and 800 ℃, respectively;
FIG. 22 is an SEM topography of uranium purified waste material, alkali treated slag, and nitric acid solution leached uranium purified waste material of example 20;
FIG. 23 is an SEM-EDS energy spectrum of uranium purified waste, alkali treated slag, and nitric acid solution leached uranium purified waste from example 20.
Detailed Description
The invention provides a method for efficiently leaching uranium and zirconium from uranium purification waste, which comprises the following steps:
(1) Mixing uranium purification waste with sodium hydroxide solution, pretreating, and separating to obtain alkali treatment slag;
(2) And (3) mixing the alkali treatment slag obtained in the step (1) with a nitric acid solution, and leaching to obtain a leaching solution containing uranium and zirconium.
The method provided by the invention is suitable for uranium purification waste containing uranium and zirconium.
In the present invention, the composition of the uranium purification scrap preferably includes Al in mass percent 2 O 3 0.25～2.25wt％，SiO 2 20.00～75.00wt％，P 2 O 5 12.50～26.29wt％，Fe 2 O 3 1.08～3.60wt％，HfO 2 0.25～0.60wt％，TiO 2 0.41～0.63wt％，K 2 O 0.35～0.41wt％，Na 2 O 0.25～1.20wt％，ZrO 2 0.12～14.40wt％，UO 2 :0.07～5wt％，CaO0.18～0.26wt％，Nb 2 O 5 0.04～0.18wt％，ThO 2 0.14～0.16wt％，SO 3 0.07 to 2.08wt% and 5 to 40wt% of an organic phase, more preferably Al 2 O 3 0.96wt％，SiO 2 26.70wt％，P 2 O 5 14.90wt％，Fe 2 O 3 1.85wt％，HfO 2 0.58wt％，TiO 2 0.63wt％，K 2 O 0.35wt％，Na 2 O 0.25wt％，ZrO 2 14.40wt％，UO 2 1.68wt％，CaO 0.25wt％，Nb 2 O 5 0.18wt％，ThO 2 0.14wt％，SO 3 0.11wt% and 37.02wt% of organic phase.
The invention mixes uranium purification waste with sodium hydroxide solution, carries out pretreatment and obtains alkali treatment slag after separation. The application can effectively pretreat by utilizing the NaOH solutionCrushing the silicon mineral component in the uranium purification waste material, dissolving the mineral matrix to produce a plurality of amorphous Zr (OH) 4 So that zirconium and SiO 2 The interlocking relationship is destroyed, thereby forming a loose porous structure, destroying the wrapped uranium phase, exposing the uranium, providing a path for leaching the uranium by the nitric acid solution, and generating Zr (OH) 4 Attached to the surface of the alkali-treated slag.
In the invention, the uranium purification waste material further comprises drying, cooling and grinding which are sequentially carried out before treatment. According to the invention, the moisture in uranium purification waste is removed through drying, and the particle size of the waste is reduced through grinding, so that the subsequent pretreatment is facilitated.
In the present invention, the temperature of the drying is preferably 105 ℃; the drying time is preferably 24 hours. The cooling operation is not particularly limited in the present invention, and may be performed by an operation well known to those skilled in the art. The operation of the grinding is not particularly limited as long as the grinding to-150 μm is ensured.
The operation of mixing the uranium purification waste material with the sodium hydroxide solution is not particularly limited in the present invention, and the technical scheme for preparing the mixed material, which is well known to those skilled in the art, may be adopted.
In the present invention, the concentration of the sodium hydroxide solution is preferably 80 to 200g/L, more preferably 100 to 180g/L, still more preferably 140g/L; the mass ratio of the volume of the sodium hydroxide solution to the uranium purification waste material is preferably (3-7) mL:1g, more preferably (4 to 5) mL:1g, more preferably 4mL:1g. The source of the sodium hydroxide solution is not particularly limited in the present invention, and may be formulated by commercially available products known to those skilled in the art or by a known formulation method. The invention further improves the leaching rate of uranium and zirconium by controlling the concentration of the sodium hydroxide solution and the mass ratio of the volume of the sodium hydroxide solution to the uranium purification waste.
In the present invention, the temperature of the pretreatment is preferably room temperature; the pretreatment time is preferably 0.5 to 3 hours, more preferably 1 to 2 hours, and even more preferably 1 hour. The invention can further improve the leaching rate of uranium and zirconium by controlling the pretreatment process parameters.
In the present invention, the pretreatment is preferably performed under stirring conditions; the rotational speed of the stirring is preferably 200 to 400rpm. The pretreatment efficiency can be further improved by carrying out pretreatment under the stirring condition.
After the pretreatment is completed, the product obtained by the pretreatment is preferably centrifuged. The centrifugation operation is not particularly limited in the present invention, and conventional operations known to those skilled in the art may be employed.
The separation operation is not particularly limited, and may be performed by an operation well known to those skilled in the art.
In the invention, the product obtained by separation is preferably washed by deionized water to obtain alkali treatment slag. In the invention, the mass ratio of the volume of deionized water to the separated product is preferably (1-13) mL:1g, more preferably (2 to 10) mL:1g, more preferably 4mL:1g. The invention can further improve the leaching rate of zirconium and uranium by limiting the volume of deionized water and the mass ratio of the separated products.
After the alkali treatment slag is obtained, the alkali treatment slag is mixed with a nitric acid solution for leaching to obtain a leaching solution containing uranium and zirconium.
The operation of mixing the alkali treatment slag with the nitric acid solution is not particularly limited, and the technical scheme for preparing the mixed material, which is well known to the person skilled in the art, can be adopted.
In the present invention, the volume ratio of nitric acid to water in the nitric acid solution is preferably (0.4 to 1.3): 1, more preferably (0.5 to 1.0): 1, more preferably 0.7:1. the source of the nitric acid solution is not particularly limited, and the nitric acid solution can be prepared by a preparation method well known to those skilled in the art. According to the invention, the volume ratio of nitric acid to water is controlled, so that the concentration of nitric acid solution is controlled, and the leaching rate of uranium and zirconium is further improved.
In the invention, the mass ratio of the volume of the nitric acid solution to the alkali treatment slag is preferably (3-7) mL:1g, more preferably (4 to 6) mL:1g, more preferably 4mL:1g. The invention can further improve the leaching rate of uranium and zirconium by controlling the volume of the nitric acid solution and the mass ratio of the alkali treatment slag.
In the present invention, the temperature of the leaching is preferably 20 to 100 ℃, more preferably 50 to 80 ℃, and even more preferably 80 ℃; the leaching time is preferably 2 to 2.5 hours. The invention can further improve the leaching rate of uranium and zirconium by controlling the leaching process parameters.
In the present invention, the leaching is preferably performed under stirring conditions; the rotational speed of the stirring is preferably 200 to 400rpm. The invention can further improve the leaching efficiency by leaching under the stirring condition.
After leaching, the invention preferably carries out centrifugation and separation on the leached product to obtain leaching liquid containing uranium and zirconium. The operation of the centrifugation and separation is not particularly limited in the present invention, and conventional operations known to those skilled in the art may be employed.
The application adopts a two-step leaching method, firstly, the silicon mineral component in uranium purification waste can be effectively crushed by pretreatment with NaOH solution, and mineral matrix is dissolved to generate a large amount of amorphous Zr (OH) 4 So that zirconium and SiO 2 The interlocking relationship is destroyed, thereby forming a loose porous structure, destroying the wrapped uranium phase, exposing the uranium, providing a path for leaching uranium by the nitric acid solution, thereby realizing the efficient leaching of uranium element and generating Zr (OH) 4 Is attached to the surface of the alkali-treated slag and further reacts with nitric acid solution in the subsequent nitric acid solution leaching process to generate ZrO (NO) 3 ) 2 The precipitate is caused to be dissolved, and the target element zirconium is dissolved in nitric acid solution to be leached out efficiently.
The method provided by the invention is to efficiently treat uranium-containing and zirconium-containing waste residues in the nuclear industry by utilizing sodium hydroxide solution pretreatment and nitric acid leaching processes, so that uranium and zirconium resources in the uranium-containing and zirconium-containing waste residues are cooperatively extracted, and the coating forms of silicon components and other valuable metal components in the waste residues are broken down by controlling sodium hydroxide pretreatment process parameters, so that target element particles are released, and high permeability and high leaching rate are obtained.
The technical solutions of the present invention will be clearly and completely described in the following in connection with the embodiments of the present invention. It will be apparent that the described embodiments are only some, but not all, embodiments of the invention. All other embodiments, which can be made by those skilled in the art based on the embodiments of the invention without making any inventive effort, are intended to be within the scope of the invention.
The uranium purification waste used in the examples was uranium-containing and zirconium-containing waste residues from Hunan province of China, which are insoluble precipitates produced during extraction filtration in certain uranium purification enterprises, and the XRF analysis results of the main components thereof are shown in Table 1 (SO in Table 1 3 Elemental sulfur is shown in the matched oxide form).
Table 1 XRF analysis results (wt.%) of the main component of uranium purification waste material
It can be seen from table 1 that the uranium purification waste material has a relatively high content of oxides such as iron and phosphorus in addition to the target elements uranium and zirconium.
NaOH、HNO 3 Are all analytically pure and purchased from national pharmaceutical chemicals, inc.
Analytical methods employed in the examples:
an X-ray fluorescence spectrometer (XRF) (S8 TIGER, germany) analyzes the chemical components of uranium and zirconium waste residues; an X-ray diffractometer (XRD) (Ultima IV, japan) analysis identified changes in the phases of as-received, alkali-treated slag and acid leaching residues; the method is characterized in that the morphological characteristics are analyzed by combining an energy dispersion X-ray Spectrum Electron Microscope (SEM) (FEI NANO450, U.S.) and an X-ray photoelectron spectroscopy (XPS) (K-Alpha, U.S.) is used for analyzing the valence state changes of uranium and zirconium before and after sample treatment, and high-precision thermal analysis (TG-DSC) (SDT Q600, U.S. TA) is used for carrying out auxiliary judgment on the thermal stability of uranium and zirconium waste residues before and after sample treatment; the concentration of the components such as U, zr in the alkaline and acidic leaching solutions was measured by inductively coupled atomic emission spectrometry (ICP) (OES 730, japan). The leaching rates (η) of U and Zr can be calculated according to formula I:
Wherein m is i The unit is g, which is the mass of elements in the leaching solution; m is m 0 The mass of elements in uranium purification waste is expressed as g; all leaching result data were obtained from the average of three replicates.
Example 1
A method for efficiently leaching uranium and zirconium from uranium purification waste, which comprises the following steps:
(1) Drying uranium purification waste materials in a 105 ℃ oven for 24 hours, taking out after natural cooling, grinding to-150 mu m, then mixing with sodium hydroxide solution, pretreating for 1 hour at room temperature, centrifuging, separating, and then washing with deionized water to obtain alkali treatment slag; wherein the concentration of the sodium hydroxide solution is 80g/L, and the mass ratio of the volume of the sodium hydroxide solution to the uranium purification waste material is 4mL:1g; the pretreatment is carried out under the stirring condition, and the stirring rotating speed is 200rpm; the mass ratio of deionized water to the separated product was 4mL:1g;
(2) Mixing the alkali treatment slag obtained in the step (1) with a nitric acid solution, leaching, centrifuging and separating to obtain a leaching solution containing uranium and zirconium; wherein the volume ratio of nitric acid to water in the nitric acid solution is 1:1, the mass ratio of the volume of the nitric acid solution to the alkali treatment slag is 4mL:1g; the leaching temperature is 80 ℃ and the leaching time is 2 hours; leaching was performed under stirring conditions at a speed of 200rpm.
Example 2
The concentration of the sodium hydroxide solution was changed to 100g/L based on example 1, and the other conditions were unchanged.
Example 3
The concentration of the sodium hydroxide solution was changed to 120g/L based on example 1, and the other conditions were unchanged.
Example 4
The concentration of the sodium hydroxide solution was changed to 140g/L based on example 1, with the other conditions unchanged.
Example 5
The concentration of the sodium hydroxide solution was changed to 160g/L based on example 1, with the other conditions unchanged.
Example 6
The concentration of the sodium hydroxide solution was changed to 180g/L on the basis of example 1, the other conditions being unchanged.
Example 7
The concentration of the sodium hydroxide solution was changed to 200g/L based on example 1, and the other conditions were unchanged.
The leaching rates and contents of uranium and zirconium in the leaching solutions of examples 1 to 7 were measured, and the results are shown in fig. 1 and 2, wherein fig. 1 shows the leaching rates and contents of uranium in the leaching solutions prepared in examples 1 to 7, and fig. 2 shows the leaching rates and contents of zirconium in the leaching solutions prepared in examples 1 to 7.
As can be seen from fig. 1 and 2, in the case of sodium hydroxide solution as an alkali pretreatment agent, the pretreatment time was 1h, and the mass ratio of the volume of the sodium hydroxide solution to the uranium purification waste material was 4mL:1g, the mass ratio of deionized water volume to the separated product is 4mL: under the condition of 1g, the influence of the concentration of sodium hydroxide solution on the extraction of uranium and zirconium is researched, when the concentration of sodium hydroxide solution is increased from 80g/L to 140g/L, the leaching rate of zirconium is gradually increased, the highest leaching rate of the zirconium can reach 93.26%, the overall leaching rate of uranium is gradually increased and then reduced, when the concentration of sodium hydroxide solution is 140g/L, the leaching rate of uranium reaches 97.96% of peak value, and meanwhile, the grade of uranium in residues is reduced to 0.06%. Because with the invasion of the high-concentration sodium hydroxide solution, the structure of uranium purification waste materials is damaged to a certain extent, micropores are generated, and the encapsulated uranium and zirconium components are released. However, with further increase of the concentration of the sodium hydroxide solution, other impurity ions such as Fe, al and the like are also released in a large amount, and excessive high-concentration sodium hydroxide solution is easier to be wasted in uranium purification SiO in the material 2 React to form Na with adhesiveness 2 SiO 3 At the same time, na in the form of silicate ion or polysilicic acid ion is present because of the increase of pH value 2 SiO 3 The more the viscosity of the solution increases, the diffusion of ions in the solution is retarded, and the leaching effect of uranium and zirconium is reduced. Thus, in combination, the optimum concentration of uranium and zirconium leached by alkali treatment is 140g/L.
Example 8
The mass ratio of sodium hydroxide solution to uranium purification waste was changed to 3mL on the basis of example 1: 1g, the other conditions are unchanged.
Example 9
The mass ratio of sodium hydroxide solution to uranium purification waste was changed to 5mL on the basis of example 1: 1g, the other conditions are unchanged.
Example 10
The mass ratio of sodium hydroxide solution to uranium purification waste was changed to 6mL on the basis of example 1: 1g, the other conditions are unchanged.
Example 11
The mass ratio of sodium hydroxide solution to uranium purification waste was changed to 7mL on the basis of example 1: 1g, the other conditions are unchanged.
The leaching rates and contents of uranium and zirconium in the leaching solutions containing uranium and zirconium in example 1 and examples 8 to 11 were measured, and the results are shown in fig. 3 and 4, wherein fig. 3 shows the leaching rates and contents of uranium in the leaching solutions prepared in example 1 and examples 8 to 11, and fig. 4 shows the leaching rates and contents of zirconium in the leaching solutions prepared in example 1 and examples 8 to 11.
As can be seen from fig. 3 and 4, at a sodium hydroxide solution concentration of 140g/L, a pretreatment time of 1h, a mass ratio of deionized water to the separated product of 4mL: under the condition of 1g, the influence of the volume of sodium hydroxide solution and the mass ratio (namely the liquid-solid ratio) of uranium purification waste (namely a sample) on uranium and zirconium extraction is studied, the leaching rate and the content of uranium and zirconium all show regular changes of increasing firstly and then suddenly decreasing along with the increase of the liquid-solid ratio, and when the liquid-solid ratio is 3:1, lower liquid-solid ratio meansTaste H in sodium hydroxide solution + Or OH (OH) - The contact with uranium purification waste is insufficient, the reaction is insufficient, the damage force and the release degree of components to the uranium purification waste are insufficient, and the leaching rate of uranium and zirconium is low; along with the increase of the liquid-solid ratio, the contact area of the sodium hydroxide solution and uranium purification waste is increased, elements such as silicon, iron, aluminum and the like are fully reacted and dissolved, and target elements such as uranium and zirconium are separated from bonding states of impurity elements, so that the uranium and zirconium are easier to participate in the subsequent nitric acid leaching process, the leaching rate is greatly improved, and the liquid-solid ratio of the sodium hydroxide solution to the uranium purification waste is 4: and 1, respectively reaching peak values of uranium leaching rate and zirconium leaching rate. But when the liquid-solid ratio is greater than 4: after 1, the extraction rate of uranium and zirconium is drastically reduced, and the injection of a large amount of sodium hydroxide solution shortens the reaction time with uranium purification waste, and simultaneously increases the leaching of impurity elements in uranium purification waste, which may react with hydroxide to generate colloid, and secondary adsorption is formed on zirconium, thereby increasing the uranium and zirconium content in the leached residues. Thus, the mass ratio of sodium hydroxide solution to uranium purification waste is optimally 4mL:1g, namely the optimal liquid-solid ratio is 4:1.
Example 12
The pretreatment time was changed to 0.5h on the basis of example 1, with the other conditions unchanged.
Example 13
The pretreatment time was changed to 1.5h on the basis of example 1, with the other conditions unchanged.
Example 14
The pretreatment time was changed to 2h on the basis of example 1, with the other conditions unchanged.
Example 15
The pretreatment time was changed to 2.5h on the basis of example 1, with the other conditions unchanged.
The leaching rates and contents of uranium and zirconium in the leaching solutions containing uranium and zirconium in example 1 and examples 12 to 15 were measured, and the results are shown in fig. 5 and 6, wherein fig. 5 shows the leaching rates and contents of uranium in the leaching solutions prepared in example 1 and examples 12 to 15, and fig. 6 shows the leaching rates and contents of zirconium in the leaching solutions prepared in example 1 and examples 12 to 15.
As can be seen from fig. 5 and 6, at a sodium hydroxide solution concentration of 140g/L, the mass ratio of sodium hydroxide solution volume to uranium purification waste material was 4mL:1g, the mass ratio of deionized water volume to the separated product is 4mL: under the condition of 1g, the influence of pretreatment time on uranium and zirconium extraction is researched, and when the pretreatment time is 1h, the leaching rates and the content of uranium and zirconium are maximized and are 97.96% and 93.26% respectively; as the pretreatment time is prolonged, the leaching rate is reduced, and when the pretreatment time is prolonged to be 2.5 hours, the leaching rate of uranium and zirconium is reduced to 91.20 percent and 63.37 percent compared with the peak value at 1 hour. This further illustrates that too long a pretreatment time may produce other side reactions, and that the released impurities may re-enter the uranium purification waste material over time, cover the particle surface, and thus affect the extraction of uranium and zirconium. Therefore, the optimal time for leaching uranium and zirconium in the pretreatment stage is 1h.
Example 16
The mass ratio of deionized water volume to the separated product was changed to 1mL based on example 1: 1g, the other conditions are unchanged.
Example 17
The mass ratio of deionized water volume to the isolated product was changed to 7mL based on example 1: 1g, the other conditions are unchanged.
Example 18
The mass ratio of deionized water volume to the separated product was changed to 10mL based on example 1: 1g, the other conditions are unchanged.
Example 19
The mass ratio of deionized water volume to the isolated product was changed to 13mL based on example 1: 1g, the other conditions are unchanged.
The leaching rates and contents of uranium and zirconium in the leaching solutions of example 1 and examples 16 to 19 were measured, and the results are shown in fig. 7 and 8, wherein fig. 7 shows the leaching rates and contents of uranium in the leaching solutions prepared in example 1 and examples 16 to 19, and fig. 8 shows the leaching rates and contents of zirconium in the leaching solutions prepared in example 1 and examples 16 to 19; the change in the pH of deionized water after the deionized water wash in example 1 and examples 16-19 is shown in fig. 9.
Because the uranium purification waste material is pretreated by adopting high-concentration alkaline solution, alkali treatment slag generated by centrifugation is strongly alkaline, alkaline components on the slag surface are firstly reacted with acid to be consumed in the subsequent nitric acid leaching process, and the leaching effect of uranium and zirconium is affected, therefore, the mass ratio of the volume of the sodium hydroxide solution to the uranium purification waste material is 4mL when the concentration of the sodium hydroxide solution is 140 g/L: 1g, under the condition that the pretreatment time is 1h, the influence of the volume of deionized water and the mass ratio of the separated products on uranium and zirconium extraction is studied, and the results are shown in figures 7 and 8.
As can be seen from fig. 7 to 9, the ratio of the volume of deionized water to the mass of the separated product has a certain influence on the leaching behavior. With the increase of the water washing degree, the leaching rate of uranium and zirconium is further improved, and the mass ratio of the volume of deionized water to the separated product is 10mL:1g (namely, the solid ratio of the water washing liquid is 10:1), when the pH value of the water washing liquid is 11.45, the leaching rates of uranium and zirconium respectively reach the peak values of 99.12 percent and 96.77 percent, and when the solid ratio of the water washing liquid is increased later, the leaching rate tends to be stable. Because the high-concentration sodium hydroxide solution erodes the uranium purification waste material in the alkaline treatment stage, the uranium purification waste material presents strong alkalinity, and the water washing liquid-solid ratio is lower than 10:1, the water washing degree is insufficient, the pH value of the solution is about 12, and the consumption of nitric acid in the subsequent acid leaching process is increased, so that the acid amount in the uranium and zirconium leaching stage is insufficient; when the solid ratio of the water washing liquid is more than 10: and 1, alkaline components floating on the surface of the alkali treatment slag are washed out, so that the contact time of direct reaction of nitric acid and uranium purification waste is increased, and the leaching rate is further improved. Therefore, after the alkaline pretreatment stage, the optimal water-wash liquor solid ratio is 10:1.
Example 20
A method for efficiently leaching uranium and zirconium from uranium purification waste, which comprises the following steps:
(1) Drying uranium purification waste materials in a 105 ℃ oven for 24 hours, taking out after natural cooling, grinding to-150 mu m, then mixing with sodium hydroxide solution, pretreating for 1 hour at room temperature, centrifuging, separating, and then washing with deionized water to obtain alkali treatment slag; wherein the concentration of the sodium hydroxide solution is 140g/L, and the mass ratio of the volume of the sodium hydroxide solution to the uranium purification waste material is 4mL:1g; the pretreatment is carried out under the stirring condition, and the stirring rotating speed is 200rpm; the mass ratio of deionized water to the separated product was 10mL:1g;
(2) Mixing the alkali treatment slag obtained in the step (1) with a nitric acid solution, leaching, centrifuging and separating to obtain a leaching solution containing uranium and zirconium; wherein the volume ratio of nitric acid to water in the nitric acid solution is 0.7:1, the mass ratio of the volume of the nitric acid solution to the alkali treatment slag is 4mL:1g; the leaching temperature is 80 ℃ and the leaching time is 2 hours; leaching was performed under stirring conditions at a speed of 200rpm.
Example 21
The volume ratio of nitric acid to water in the nitric acid solution was changed to 0.4 based on example 20: 1, the other conditions are unchanged.
Example 22
The volume ratio of nitric acid to water in the nitric acid solution is changed to 1 on the basis of the embodiment 20: 1, the other conditions are unchanged.
Example 23
The volume ratio of nitric acid to water in the nitric acid solution was changed to 1.3 on the basis of example 20: 1, the other conditions are unchanged.
Comparative example 1
The volume ratio of nitric acid to water in the nitric acid solution was changed to 1.6 on the basis of example 20: 1, the other conditions are unchanged.
The leaching rates and contents of uranium and zirconium in the leaching solutions of examples 20 to 23 and comparative example 1 were measured, and the results are shown in fig. 10 and 11, wherein fig. 10 shows the leaching rates and contents of uranium in the leaching solutions prepared in examples 20 to 23 and comparative example 1, and fig. 11 shows the leaching rates and contents of zirconium in the leaching solutions prepared in examples 20 to 23 and comparative example 1.
As can be seen from fig. 10 and 11, in the pretreatment process of sodium hydroxide solution, components such as iron and aluminum adhere to the surface of uranium purification waste in the formation of precipitate under the action of strong alkali, and as the concentration of nitric acid solution increases continuously, the leaching rate and content of uranium and zirconium show a small-amplitude rising trend, and the volume ratio is 0.7:1, the leaching rates of uranium and zirconium are 99.37 percent and 97.48 percent respectively; then a drop starts to occur until the volume ratio is 1.6: at 1, the leaching rate suddenly drops. The result shows that the higher the concentration of the acid is, the better is the leaching process, and as the concentration of the acid is increased, unnecessary impurity elements are leached instead, thereby generating side reactions, and leading to the reduction of leaching rate. Thus, the optimal solution in the nitric acid leaching process is 0.7 by volume: 1 in nitric acid.
Example 24
The mass ratio of the nitric acid solution to the alkali treatment slag was changed to 3mL on the basis of example 20: 1g, the other conditions are unchanged.
Example 25
The mass ratio of the nitric acid solution to the alkali treatment slag was changed to 5mL on the basis of example 20: 1g, the other conditions are unchanged.
Example 26
The mass ratio of the volume of the nitric acid solution to the alkali treatment slag was changed to 6mL on the basis of example 20: 1g, the other conditions are unchanged.
Comparative example 2
The mass ratio of the nitric acid solution to the alkali treatment slag was changed to 2mL on the basis of example 20: 1g, the other conditions are unchanged.
The leaching rates and contents of uranium and zirconium in the leaching solutions of examples 20, examples 24 to 26 and comparative example 2 were measured, and the results are shown in fig. 12 and 13, wherein fig. 12 shows the leaching rates and contents of uranium in the leaching solutions prepared in examples 20, examples 24 to 26 and comparative example 2, and fig. 13 shows the leaching rates and contents of zirconium in the leaching solutions prepared in examples 20, examples 24 to 26 and comparative example 2.
As can be seen from fig. 12 and 13, the leaching rate and content of uranium and zirconium are related to the mass ratio of the nitric acid solution to the alkali treated slag (i.e., HNO 3 The solid ratio of the leaching solution) increases, the trend of increasing and then decreasing is shown, and the solid ratio of the leaching solution is 4:1, the uranium and zirconium reach leaching peaks Values. When the liquid-solid ratio is further increased, the leaching effect of uranium and zirconium is reduced, so that the optimal liquid-solid ratio in the nitric acid leaching process is 4:1.
example 27
The leaching temperature was changed to 20 ℃ on the basis of example 20, with the other conditions unchanged.
Example 28
The leaching temperature was changed to 40 ℃ on the basis of example 20, with the other conditions unchanged.
Example 29
The leaching temperature was changed to 60 ℃ on the basis of example 20, with the other conditions unchanged.
Example 30
The leaching temperature was changed to 100 ℃ on the basis of example 20, with the other conditions unchanged.
The leaching rates and contents of uranium and zirconium in the leaching solutions of example 20 and examples 27 to 30 were measured, and the results are shown in fig. 14 and 15, wherein fig. 14 shows the leaching rates and contents of uranium in the leaching solutions prepared in example 20 and examples 27 to 30, and fig. 15 shows the leaching rates and contents of zirconium in the leaching solutions prepared in example 20 and examples 27 to 30.
As can be seen from figures 14 and 15, at low temperatures, the uranium and zirconium leaches to a lesser extent, and as the temperature increases, energy is deposited in the porous structure, creating a temperature gradient between the solid and liquid phases, and a greater thermal convection diffusion through the temperature gradient in the solution, causing the reaction to proceed in a direction favoring uranium zirconium leaching and peaking at 80 ℃. When the temperature reaches 100 ℃, the solution evaporation can be caused to influence the leaching effect of nitric acid on uranium and zirconium. Thus, the optimum temperature for nitric acid leaching is 80 ℃.
Physicochemical evolution of uranium purification waste in a two-step leaching process
Dynamic evolution of uranium purification waste element content and target element valence state in (one) two-step leaching process
In the leaching process of uranium and zirconium, the impurity elements such as iron, silicon, phosphorus and the like are important influencing factors for influencing the extraction of uranium and zirconium on the wrapping of uranium and zirconium and the side reaction generated in the treatment stage of uranium purification waste. Elemental analysis was therefore performed on the uranium purified waste material from the two-step leaching process of example 20, the results being shown in table 2.
Table 2 results of ICP-MS analysis of major elements of uranium purification waste in a two-step leaching process (wt.%)
The XPS analysis results of uranium and zirconium in the two-step leaching process of example 20 are shown in fig. 16 to 18, wherein fig. 16 is a carbon total spectrum of the uranium purified waste material, the alkali treated slag, and the uranium purified waste material after nitric acid solution leaching in example 20; FIG. 17 is a uranium spectrum of uranium purified waste material, alkali treated slag, and nitric acid solution leached uranium purified waste material of example 20; FIG. 18 is a zirconium chart of uranium purified waste, alkali treated slag and uranium purified waste after nitric acid solution leaching in example 20 (in the figure, the original curve is marked, namely, the curve of uranium purified waste, the curve of NaOH is marked, namely, the curve of alkali treated slag, and the curve of NaOH-HNO is marked) 3 I.e. the curve of uranium purified waste after nitric acid solution leaching, the same applies below).
As can be seen from fig. 16 to 18, the target elements uranium and zirconium are not leached during the alkali treatment, but rather enrichment occurs with a decrease in the amount of uranium purification waste. And the contents of impurity elements of aluminum, iron, sodium and chlorine are increased to different degrees, and only phosphorus and silicon elements are leached in a large amount in the alkaline treatment stage. Because the structure of the uranium purification waste is damaged to a certain extent along with the invasion of the high-concentration NaOH solution, the components such as silicate, phosphate and the like in the uranium purification waste are decomposed, and therefore the dissociation of uranium is promoted. At the same time, iron in the uranium purification waste material is dissociated along with the reaction, wherein Fe 3+ Can be used as an oxidizing agent of uranium to convert insoluble U (IV) into soluble U (VI). Further characterization analysis combined with X-ray photoelectron spectroscopy (XPS) can see the change of the electronic structure of uraniumAnd (5) melting. From the original sample (i.e., uranium purification waste) to the alkali treatment, U (IV) was largely converted to U (VI) and XPS spectra peaks were not shifted, it was speculated that uranium in the alkali treatment did not react with other elements to produce new species. The valence state of zirconium is always +4, but the peak of zirconium in uranium purification waste after alkali treatment is shifted, which may be caused by further chemical reaction of zirconium or partial electron transfer given by impurity elements of uranium purification waste in the process. In the subsequent nitric acid leaching process, the target elements uranium and zirconium are leached in large quantities, with optimal leaching rates of 99.37% and 97.48%, respectively, because both uranium and zirconium are converted into phases susceptible to leaching by acid in the alkaline treatment stage. Meanwhile, the uranium content in uranium purified waste after acid leaching is too low and no corresponding uranium valence state characteristic peak appears, so that fitting is not possible, and the result is identical with the uranium content result of ICP analysis in Table 2; the peak fitted by zirconium is still in the range of +4 valence after alkali treatment, and is overlapped with the peak of the initial uranium purification waste after nitric acid leaching, and the combination of the change condition of increasing and decreasing the content of zirconium element in the two-step leaching process and the analysis of 1.14% of zirconium residues in the uranium purification waste after nitric acid leaching is still carried out, probably because most of zirconium is converted into a phase which is easy to be leached by later nitric acid in the alkali treatment process, the phase is covered on the surface of the uranium purification waste, and a small amount of unconverted part is still tightly combined with the uranium purification waste and covered by a new conversion phase. The subsequent nitric acid leaching stage dissolves the easily extracted zirconium components attached to the surface in nitric acid solution, and therefore a small amount of unconverted zirconium components in the uranium purification waste are exposed, so that the change condition that the valence peaks of zirconium in the two-step leaching process are offset and coincide again in the figure is formed. And finally two peaks at 183.12eV and 185.50eV, zr3 respectively d5/2 And Zr3 d3/2 It has also been shown that the morphology of the leached zirconium in the two-stage leaching process is only Zr (IV).
Phase evolution of uranium purification waste in (two) two-step leaching process
The XRD diffractogram of the uranium purified waste material, alkali treated slag, and uranium purified waste material (i.e., residue) after nitric acid solution leaching in example 20 is shown in fig. 19; the TG-DEC curve of the alkali treated slag in example 20 is shown in fig. 20; the XRD diffraction patterns of the alkali-treated slag in example 20 after roasting at 300 ℃, 600 ℃ and 800 ℃ respectively are shown in FIG. 21.
As can be seen from FIG. 19, siO in the residue, which is uranium purification waste material after alkali treatment and re-nitric acid leaching 2 Characteristic peaks gradually stand out, peak intensities increase, while the base line of the diffraction pattern overall gradually goes to a level, and the impurity peaks obviously decrease. The reaction in the two-step leaching process is combined, and the whole process is actually the process of impurity removal, purification and extraction of uranium purification waste, which is compatible with the change of the content of non-target elements in table 2. After the alkali treatment, the uranium purification waste is decomposed, and phosphorus element therein is leached in a large amount, and the characteristic peak intensities of corresponding phosphorus at 2θ=8.82° and 2θ=19.94° are reduced in the figure. The dissolution of uranium purification waste and the stripping of impurity elements enable the uranium as a target element to be in full contact with nitric acid solution, and the U (VI) which is partially insoluble in the valence state of uranium after alkali treatment in the combination of figures 16-18 is converted into the U (VI) which is easily soluble, so that the uranium extraction rate is greatly improved to a certain extent.
However, in the case of zirconium, the element content of the alkali-treated slag was enriched from the initial 13.33% to 20.20% according to Table 2, but the diffraction pattern of the alkali-treated slag in FIG. 19 did not find a characteristic peak matching with the alkali-treated slag, and the diffraction peak was widened in a part of the area, probably because the crystallinity of the zirconium product after the alkali treatment was low, and could not be shown in the XRD pattern. Thus, the alkali-treated slag was sampled for TG-DSC analysis. As can be seen from FIG. 20, at a heating rate of 10 ℃/min, the TG curve is always in the falling phase, due on the one hand to the water content of the alkali-treated slag and on the other hand to the thermal loss of the volatile substances. At the same time, the phenomena of heat absorption and weight loss occur between 0 ℃ and 100 ℃, and an exothermic interval occurs between 300 ℃ and 800 ℃, but no corresponding weight loss occurs on the corresponding DTG curve, and the process is that the substance crystal form is transformed. To study the phase change process and the crystal form change, test pairsThe alkali-treated slag was calcined at 300, 600, and 800 ℃ for 5 hours, respectively, and then subjected to XRD scanning for analyte phases. As can be seen from FIG. 21, there is tetragonal ZrO in the calcined product at 300 ℃ 2 Presence; after calcination at 600 ℃, zrO with tetragonal phase 2 And also has monoclinic phase ZrO 2 The presence of tetragonal ZrO even under calcination at 800 DEG C 2 Exists. This rule is associated with Zr (OH) 4 Is dehydrated and decomposed to form ZrO 2 The crystal form transformation rules are identical, and in combination with FIG. 20, the endothermic peak at 0-100 ℃ should be Zr (OH) 4 Decomposition into ZrO 2 The reaction is shown in the formula II, and the exothermic process at 300-800 ℃ is ZrO 2 A crystal form transformation occurs. This is also explained by the formation of amorphous Zr (OH) 4 The reason why the alkali treated slag does not have a characteristic peak of zirconium to stand out in XRD scanning in fig. 19.
Zr(OH) 4 (s)→ZrO 2 (s)+2H 2 O (g) formula II
Zr (OH) generated after alkali treatment 4 Medium zirconium is +4 valent, while ZrP in raw uranium purified waste material 2 O 7 The zirconium in the solution is +4, which is consistent with the change of the valence state of the zirconium element in FIGS. 16-18, which also proves that ZrP is present during alkali treatment 2 O 7 Reaction with NaOH produces Zr (OH) in large amounts that is readily leached by nitric acid 4 And covers the surface of uranium purification waste material, so that a small amount of valence peak shift of zirconium element after alkali treatment in FIGS. 16-18 is formed, and finally Zr (OH) 4 Is dissolved and extracted in the subsequent nitric acid leaching process, and only a small amount of non-converted ZrP is left in the residue 2 O 7 In fig. 16 to 18, the zirconium valence peak is shown to be in the initial position. In summary, it can be seen that the crystal phase of the zirconium component in the uranium purification waste is not stable, but is convenient for subsequent extraction of zirconium.
Microcosmic morphology change of uranium purification waste in (three) two-step leaching process
SEM morphology of the uranium purified waste material, the alkali treated slag and the nitric acid solution leached uranium purified waste material of example 20 is shown in fig. 22.
As can be seen from fig. 22, the as-received uranium purification waste material has a relatively complete structure, is compact and is agglomerated, and is unfavorable for the sufficient contact between the leachate and the target element, resulting in low reaction efficiency and poor leaching effect. After being treated by NaOH solution, the uranium purification waste material has a broken surface structure, becomes loose, increases the specific surface area, further reduces the particle size and distributes on the surface in fine particles, so that the reaction in the subsequent nitric acid leaching stage can be more fully carried out. The micro-pore structure can be clearly observed after the uranium purification waste is leached by nitric acid, which proves that the alkali treatment can indeed damage the element embedded structure in the uranium purification waste, thereby providing a sufficient contact opportunity for the reaction of zirconium extraction and providing a path for the enhanced leaching of uranium.
The SEM-EDS energy spectrum of the uranium purified waste material, the alkali treated slag and the nitric acid solution leached uranium purified waste material in example 20 is shown in fig. 23.
As can be seen from fig. 23, after the uranium purification waste is subjected to alkali treatment, the bulk structure is changed into a loose and porous structure, and the content of phosphorus is greatly reduced; unlike the enrichment of uranium and zirconium as target elements in the data of table 2 in the alkaline treatment stage, the reduction phenomenon also occurs in the uranium and zirconium contents in the figure. In combination with the characteristic peaks of silicon and oxygen in fig. 23, the main component of the scanned area is basically quartz, and the alkali treatment process is equivalent to stripping uranium and zirconium from the structure, leaving insoluble quartz components and finally presenting a loose and porous shape, which also explains the problem of the reduction of the characteristic peaks of the energy spectra of uranium and zirconium in the graph. In the acid leaching process, the contents of iron, phosphorus and aluminum in the residues are further reduced, and the leaching rates of uranium and zirconium are obviously increased, so that the alkaline solution damages the structure of uranium purification waste to a certain extent, the formed porous structure is favorable for subsequent nitric acid leaching, and a leaching channel is provided for the subsequent nitric acid solution to be immersed into the uranium purification waste.
From the above results, it can be seen that the uranium and the silicon mineral components in the uranium purification waste material have symbiotic or interlocking relationship, and the pretreatment by NaOH solution can effectively crush the silicon mineral components, dissolve mineral matrix, destroy the uranium phase by the structure of the encapsulated uranium purification waste material, and expose the uranium. And the dissolved uranium purification waste presents a loose porous structure, and the micropores also provide a path for leaching uranium by nitric acid. In addition, dissolution of the uranium purification waste material results in an increase in specific surface area such that the uranium purification waste material is in contact with H in a subsequent nitric acid leaching process + The reaction can be more sufficiently conducted.
In the alkaline treatment process, in order to obtain a higher uranium leaching rate, the concentration of the alkali needs to be increased as much as possible, so that the contents of impurity elements Si, P and the like in uranium purification waste are reduced (table 2), silicate and phosphate components in the uranium purification waste are effectively dissolved when the concentration of NaOH solution is 140g/L in combination with the above-mentioned condition factors, the uranium purification waste is destroyed to the greatest extent, the combination of the impurity elements and NaOH is reduced, the solution viscosity is increased, the diffusion of ions in the solution is blocked, the aperture of the dissolved out aperture is reduced, and the reduced particle size of the uranium purification waste also promotes the rapid and efficient completion of uranium leaching, which is far higher than the uranium leaching rate of single acid leaching.
Zirconium is mainly prepared from unstable crystal form substance ZrP in original uranium purification waste material 2 O 7 In existence, the uranium purification waste material after NaOH treatment disintegrates in structure to generate a large amount of amorphous Zr (OH) 4 So that zirconium and SiO 2 The interlocking relationship between them is broken so that the dissolution forms a loose porous structure. And dissolve the generated Zr (OH) 4 Attached to the surface of uranium purification waste material, and further reacted with nitric acid solution in a subsequent nitric acid leaching process to produce ZrO (NO) 3 ) 2 The precipitate is dissolved, and the target element zirconium is also dissolved in nitric acid solution to be extracted, and the specific reaction is shown in formulas III and IV:
ZrP 2 O 7 +4NaOH→Zr(OH) 4 ↓+Na 4 P 2 O 7 formula III;
Zr(OH) 4 +2HNO 3 →ZrO(NO 3 ) 2 +3H 2 o formula IV.
As can be seen from the above examples, the method provided by the present invention pretreats-HNO by using NaOH solution 3 The leaching technology is used for efficiently treating uranium-containing and zirconium-containing waste residues in uranium purification industry so as to improve the content of impurity elements in the waste residues and uranium purification waste materialsThe structure increases the uranium and zirconium extraction effect, thereby realizing the efficient collaborative extraction of uranium and zirconium resources and finally realizing the remarkable achievements of 99.37 percent of uranium leaching rate and 97.48 percent of zirconium leaching rate. The whole process does not introduce other impurities, and the process is simple and convenient to operate, thereby providing technical support for clean and efficient treatment of the valuable metal-containing solid waste.
The foregoing is merely a preferred embodiment of the present invention and it should be noted that modifications and adaptations to those skilled in the art may be made without departing from the principles of the present invention, which are intended to be comprehended within the scope of the present invention.
Claims (10)
1. A method for efficiently leaching uranium and zirconium from uranium purification waste, comprising the steps of:
(1) Mixing uranium purification waste with sodium hydroxide solution, pretreating, and separating to obtain alkali treatment slag;
(2) And (3) mixing the alkali treatment slag obtained in the step (1) with a nitric acid solution, and leaching to obtain a leaching solution containing uranium and zirconium.
2. The method according to claim 1, wherein the concentration of the sodium hydroxide solution in the step (1) is 80-200 g/L, and the mass ratio of the volume of the sodium hydroxide solution to the uranium purification waste material is (3-7) mL:1g.
3. The method according to claim 1, wherein the temperature of the pretreatment in the step (1) is room temperature, and the pretreatment time is 0.5 to 3 hours.
4. A method according to claim 1 or 3, wherein the pretreatment in step (1) is carried out under stirring conditions at a rotational speed of 200 to 400rpm.
5. The method of claim 1, wherein the product isolated in step (1) is subjected to deionized water washing.
6. The method according to claim 5, wherein the mass ratio of the volume of deionized water to the separated product is (1-13) mL:1g.
7. The method of claim 1, wherein the volume ratio of nitric acid to water in the nitric acid solution in step (2) is (0.4-1.3): 1.
8. the method according to claim 7, wherein the mass ratio of the volume of the nitric acid solution to the alkali treatment slag in the step (2) is (3-7) mL:1g.
9. The method according to claim 1, wherein the leaching in step (2) is carried out at a temperature of 20 to 100 ℃ for a time of 2 to 2.5 hours.
10. The method according to claim 1 or 9, wherein the leaching temperature in step (2) is 50-80 ℃.
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