CN111826521B - Method for removing and separating uranium and thorium from ferrocolumbium concentrate - Google Patents

Method for removing and separating uranium and thorium from ferrocolumbium concentrate Download PDF

Info

Publication number
CN111826521B
CN111826521B CN202010729164.3A CN202010729164A CN111826521B CN 111826521 B CN111826521 B CN 111826521B CN 202010729164 A CN202010729164 A CN 202010729164A CN 111826521 B CN111826521 B CN 111826521B
Authority
CN
China
Prior art keywords
leaching
thorium
uranium
concentrate
rate
Prior art date
Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
Active
Application number
CN202010729164.3A
Other languages
Chinese (zh)
Other versions
CN111826521A (en
Inventor
李江
柯平超
刘亚洁
刘超
王学刚
孙占学
陈一鸣
胡艺杰
崔文馨
解原
魏嘉欣
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
GUANGDONG INSTITUTE OF MINERAL APPLICATION
East China Institute of Technology
Original Assignee
Guangdong Institute Of Mineral Application
East China Institute of Technology
Priority date (The priority date is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the date listed.)
Filing date
Publication date
Application filed by Guangdong Institute Of Mineral Application, East China Institute of Technology filed Critical Guangdong Institute Of Mineral Application
Priority to CN202010729164.3A priority Critical patent/CN111826521B/en
Publication of CN111826521A publication Critical patent/CN111826521A/en
Application granted granted Critical
Publication of CN111826521B publication Critical patent/CN111826521B/en
Active legal-status Critical Current
Anticipated expiration legal-status Critical

Links

Images

Classifications

    • 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
    • C22B3/00—Extraction of metal compounds from ores or concentrates by wet processes
    • C22B3/04—Extraction of metal compounds from ores or concentrates by wet processes by leaching
    • C22B3/06—Extraction of metal compounds from ores or concentrates by wet processes by leaching in inorganic acid solutions, e.g. with acids generated in situ; in inorganic salt solutions other than ammonium salt solutions
    • C22B3/10—Hydrochloric acid, other halogenated acids or salts thereof
    • 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
    • C22B3/00—Extraction of metal compounds from ores or concentrates by wet processes
    • C22B3/04—Extraction of metal compounds from ores or concentrates by wet processes by leaching
    • C22B3/12—Extraction of metal compounds from ores or concentrates by wet processes by leaching in inorganic alkaline solutions
    • C22B3/14—Extraction of metal compounds from ores or concentrates by wet processes by leaching in inorganic alkaline solutions containing ammonia or ammonium salts
    • 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
    • C22B60/023—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching using acidic solutions or liquors halogenated ion as active agent
    • 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
    • C22B60/0291—Obtaining thorium, uranium, or other actinides obtaining thorium
    • 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

Landscapes

  • Chemical & Material Sciences (AREA)
  • Engineering & Computer Science (AREA)
  • General Life Sciences & Earth Sciences (AREA)
  • Life Sciences & Earth Sciences (AREA)
  • Geology (AREA)
  • Manufacturing & Machinery (AREA)
  • Environmental & Geological Engineering (AREA)
  • Materials Engineering (AREA)
  • Mechanical Engineering (AREA)
  • Metallurgy (AREA)
  • Organic Chemistry (AREA)
  • Inorganic Chemistry (AREA)
  • Geochemistry & Mineralogy (AREA)
  • Manufacture And Refinement Of Metals (AREA)
  • Inorganic Compounds Of Heavy Metals (AREA)

Abstract

The invention provides a method for removing and separating uranium and thorium from ferrocolumbium concentrate, which comprises the following steps: leaching uranium in the ferrocolumbium concentrate in multiple stages by using a leaching agent containing acidic fluoride ions to obtain a uranium-containing leachate and leaching slag; and (3) carrying out multistage leaching on thorium in the leaching slag by using an ammonium salt solution to obtain a thorium-containing leaching solution and ferrocolumbium concentrate from which uranium and thorium are removed. According to the method, the niobium-iron ore concentrate is leached by using the leaching agent containing acidic fluoride ions, the surfaces and the gap interfaces of niobium-iron ore concentrate particles are slowly corroded and dissolved by using the strong corrosivity of an acidic fluoride ion solution, the specific surface area of the ore particles is increased, and the state of the particle interfaces is activated, so that the leaching of uranium is promoted; meanwhile, thorium in the ore is converted into stable ThF by utilizing the strong coordination capacity of fluorine ions and thorium ions4Precipitating and separating from uranium-containing leaching solution; leaching the leaching slag by using ammonium salt solution, and leaching ThF by using ammonium ions4The strong solubility of the thorium can realize the selective leaching removal of the thorium.

Description

Method for removing and separating uranium and thorium from ferrocolumbium concentrate
Technical Field
The invention belongs to the technical field of hydrometallurgy, and particularly relates to a method for removing and separating uranium and thorium from ferrocolumbium concentrate.
Background
Niobium has the characteristics of good superconductivity, corrosion resistance, wear resistance and the like, is widely applied to the fields of steel, superconducting materials, aerospace, atomic energy and the like, and is one of key materials for development of high and new technologies such as information technology, new energy technology, space technology, biotechnology, superconducting technology and the like. The distribution of the world niobium resources is extremely uneven, and brazil alone accounts for more than 95% of the total reserves in the world. China is a big country lacking niobium, 98% of China is imported, and in all imported countries, the columbite which is imported from Brazil only accounts for 95% of the total amount. The main niobium-containing ore phase in the Brazilian ferrocolumbium concentrate is pyrochlore, however, the pyrochlore ore phase contains a certain amount of radioactive elements of uranium and thorium, and serious potential safety hazards exist.
In the prior art, the removal of radioactive elements such as uranium, thorium and the like from ferrocolumbium concentrate generally adopts an acid leaching method and an alkaline leaching method. The existing known acid leaching method is to recover the radioactive elements in the niobium ore by adopting a four-acid mixing method (hydrochloric acid, nitric acid, hydrofluoric acid and perchloric acid); the alkaline leaching method is a normal-pressure alkaline stirring leaching method, and the optimal leaching conditions are as follows: the liquid-solid ratio is 1.2:1, the temperature is 70-75 ℃, the leaching time is 4 hours, and the residual alkali is 85g/L (Du Ying. research on leaching mechanism of radioactive element uranium-thorium in the high-fluorine strong acid tantalum-niobium slag of a certain place [ D ] 2019 ]. However, because uranium and thorium mostly exist in the form of similar images, the two leaching processes can only remove radioactive elements from ferrocolumbium concentrate to obtain the radioactive elements mixed together, and uranium and thorium in the radioactive elements cannot be separated.
Based on this, a need exists for a method for removing and separating uranium and thorium from ferrocolumbium concentrates.
Disclosure of Invention
The invention aims to provide a method for removing and separating uranium and thorium from ferrocolumbium concentrate. The method provided by the invention can remove uranium and thorium from the ferrocolumbium concentrate and separate the uranium and the thorium.
In order to achieve the above object, the present invention provides the following technical solutions:
the invention provides a method for removing and separating uranium and thorium from ferrocolumbium concentrate, which comprises the following steps:
(1) leaching uranium in the ferrocolumbium concentrate in multiple stages by using a leaching agent containing acidic fluoride ions to obtain a uranium-containing leachate and leaching slag;
(2) and (2) carrying out multistage leaching on thorium in the leaching slag obtained in the step (1) by using an ammonium salt solution to obtain a thorium-containing leaching solution and niobium-iron concentrate subjected to uranium and thorium removal.
Preferably, the leaching agent containing acidic fluoride ions in the step (1) is a mixed acid system, and the mixed acid system is HF-H2SO4、HF-H2O2、HF-H2SO4-H2O2HF-HCl and HF-HCl-H2O2One kind of (1).
Preferably, the concentration of the fluoride ions in the leaching agent containing acidic fluoride ions in the step (1) is 10-50 g/L.
Preferably, the temperature of the multistage leaching in the step (1) and the temperature of the multistage leaching in the step (2) are 60-95 ℃ independently.
Preferably, the number of stages of the multistage leaching in the step (1) and the multistage leaching in the step (2) is independently two or more stages.
Preferably, the time of each leaching stage in the multi-stage leaching in the step (1) is independently 3-8 h, and the time of each leaching stage in the multi-stage leaching in the step (2) is independently 20-120 min.
Preferably, the volume ratio of the mass of the ferrocolumbium concentrate in the step (1) to the leaching agent containing acidic fluoride ions is 1 g: (5-30) mL.
Preferably, the ammonium salt in the ammonium salt solution in the step (2) includes at least one of ammonium carbonate, ammonium bicarbonate, ammonium sulfate, ammonium bisulfate, ammonium sulfite and ammonium chloride.
Preferably, NH in the ammonium salt solution in the step (2)4 +The concentration of (b) is 0.5 to 3.0 mol/L.
Preferably, the volume ratio of the mass of the leaching residue to the ammonium salt solution in the step (2) is 1 g: (5-30) mL.
The invention provides a method for removing and separating uranium and thorium from ferrocolumbium concentrate, which comprises the following steps: leaching uranium in the ferrocolumbium concentrate in multiple stages by using a leaching agent containing acidic fluoride ions to obtain a uranium-containing leachate and leaching slag; and (3) carrying out multistage leaching on thorium in the leaching slag by using an ammonium salt solution to obtain a thorium-containing leaching solution and ferrocolumbium concentrate from which uranium and thorium are removed. According to the method, the niobium-iron ore concentrate is leached by using the leaching agent containing acidic fluoride ions, the surfaces and the gap interfaces of niobium-iron ore concentrate particles are slowly corroded and dissolved by using the strong corrosivity of an acidic fluoride ion solution, the specific surface area of the ore particles is increased, and the state of the particle interfaces is activated, so that the leaching of uranium is promoted; meanwhile, thorium in the ore is converted into stable ThF by utilizing the strong coordination capacity of fluorine ions and thorium ions4Precipitating and separating from uranium-containing leaching solution; leaching the leaching slag by using ammonium salt solution, and leaching ThF by using ammonium ions4The strong solubility of the thorium can realize the selective leaching removal of the thorium. The results of the examples show that after the ferrocolumbium concentrate is treated by the method provided by the invention, the removal rates of uranium and thorium in the ferrocolumbium concentrate are both higher than 90%, and the separation rates are both higher than 90%.
Drawings
FIG. 1 is a plot of the uranium and thorium leaching rates as a function of leaching time for a two-stage uranium leaching process of example 1;
FIG. 2 is a plot of the uranium and thorium leaching rates as a function of leaching time during the three-stage leaching of the thorium of example 1.
Detailed Description
The invention provides a method for removing and separating uranium and thorium from ferrocolumbium concentrate, which comprises the following steps:
(1) leaching uranium in the ferrocolumbium concentrate in multiple stages by using a leaching agent containing acidic fluoride ions to obtain a uranium-containing leachate and leaching slag;
(2) and (2) carrying out multistage leaching on thorium in the leaching slag obtained in the step (1) by using an ammonium salt solution to obtain a thorium-containing leaching solution and niobium-iron concentrate subjected to uranium and thorium removal.
The method provided by the invention is suitable for the ferrocolumbium concentrate containing radioactive elements of uranium and thorium. In the present invention, the ferrocolumbium concentrate is preferably brazilian radioactive ferrocolumbium concentrate. In the present invention, the main mineral phase of the brazilian radioactive ferrocolumbium concentrate preferably includes at least one of pyrochlore, ilmenite, monazite and micanite.
In the invention, the particle size of the ferrocolumbium concentrate is preferably 10-80 μm, more preferably 20-60 μm, and even more preferably 30-50 μm. In the invention, when the particle size of the ferrocolumbium concentrate is in the range, the acidic fluoride ion solution can be promoted to corrode and dissolve the surface and the gap interface of the ferrocolumbium concentrate particles, so that the specific surface area of the ore particles is increased, the effect of activating the state of the particle interface is achieved, and the leaching rate of uranium is further improved. In the present invention, when the particle size of the ferrocolumbium concentrate does not satisfy the above conditions, it is preferable that the ferrocolumbium concentrate is crushed first. The crushing mode of the ferrocolumbium concentrate is not particularly limited, and the crushing mode known by the technical personnel in the field can be adopted.
The invention uses the leaching agent containing acidic fluoride ions to carry out multi-stage leaching on uranium in the ferrocolumbium concentrate to obtain uranium-containing leachate and leaching slag. In the invention, when the leaching agent containing the acidic fluoride ions is used for leaching the ferrocolumbium concentrate, the strong corrosivity of the acidic fluoride ion solution is utilized to slowly corrode and dissolve the surfaces and the gap interfaces of the ferrocolumbium concentrate particles, so that the specific surface area of the ore particles is increased, and the state of the particle interfaces is activated, thereby promoting the leaching of uranium; meanwhile, thorium in the ore is converted into stable ThF by utilizing the strong coordination capacity of fluorine ions and thorium ions4Precipitating and separating from the uranium-containing leaching solution.
In the invention, the leaching agent containing acidic fluoride ions is preferably a mixed acid system, and the mixed acid system is preferably HF-H2SO4、HF-H2O2、HF-H2SO4-H2O2HF-HCl and HF-HCl-H2O2One kind of (1). In the present invention, the leaching agent containing acidic fluoride ions preferably further includes water. The dosage of the water is not specially limited, and the components in the leaching agent containing the acidic fluoride ions can meet the concentration requirement.
In the invention, the concentration of the fluoride ions in the leaching agent containing the acidic fluoride ions is preferably 10-50 g/L, more preferably 20-45 g/L, and even more preferably 30-40 g/L. In the invention, when the concentration of the fluorine ions is in the range, the strong corrosivity of the acidic fluorine ion solution can further corrode and dissolve the surface and the gap interface of the ferrocolumbium concentrate particles, increase the specific surface area of the ore particles and activate the interface state of the particles, thereby further promoting the leaching of uranium; meanwhile, thorium in the ore is converted into stable ThF by utilizing the strong coordination capacity of fluorine ions and thorium ions4Precipitating and separating from the uranium-containing leaching solution.
In the invention, when the mixed acid system is HF-H2SO4In the leaching agent containing the acidic fluoride ions, SO is4 2-The concentration of (b) is preferably 30 to 100g/L, more preferably 40 to 80g/L, and still more preferably 60 to 70 g/L. In the invention, the mixed acid system is HF-H2SO4In the process, the strong corrosivity of the acid fluoride ions is utilized to slowly corrode and dissolve the surfaces of the niobium-iron ore concentrate particles and the gap interfaces, so that the specific surface area of the ore particles is increased, and the state of the particle interfaces is activated, thereby promoting the leaching of uranium; meanwhile, thorium in the ore is converted into stable ThF by utilizing the strong coordination capacity of fluorine ions and thorium ions4Precipitating and separating from uranium-containing leaching solution; said H2SO4And an acid environment is provided for dissolving the oxidizing gangue components K, Mg, Al, U, Th and the like in the ferrocolumbium concentrate.
In the invention, the mixed acid system is HF-H2O2In the meantime, H is used for preparing the leaching solution2O2Meter, said H2O2The concentration of (b) is preferably 10 to 30g/L, more preferably 15 to 30g/L, and still more preferably 20 to 25 g/L. In the invention, the mixed acid system is HF-H2O2When usingThe strong corrosivity of the acid fluoride ions slowly erodes and dissolves the surfaces and the gap interfaces of the ferrocolumbium concentrate particles, the specific surface area of the ore particles is increased, and the state of the particle interfaces is activated, so that the leaching of uranium is promoted; meanwhile, thorium in the ore is converted into stable ThF by utilizing the strong coordination capacity of fluorine ions and thorium ions4Precipitating and separating from uranium-containing leaching solution; said H2O2The oxidability of the leaching solution is improved, and the oxidation leaching of low-valence uranium in the ferrocolumbium concentrate is promoted.
In the invention, the mixed acid system is HF-H2SO4-H2O2In the leaching agent containing the acidic fluoride ions, SO is4 2-The concentration of (b) is preferably 30-100 g/L, more preferably 40-80 g/L, and even more preferably 60-70 g/L; h used for preparing leachate2O2Meter, said H2O2The concentration of (b) is preferably 10 to 30g/L, more preferably 15 to 30g/L, and still more preferably 20 to 25 g/L. In the invention, the mixed acid system is HF-H2SO4-H2O2In the process, the strong corrosivity of the acid fluoride ions is utilized to slowly corrode and dissolve the surfaces of the niobium-iron ore concentrate particles and the gap interfaces, so that the specific surface area of the ore particles is increased, and the state of the particle interfaces is activated, thereby promoting the leaching of uranium; meanwhile, thorium in the ore is converted into stable ThF by utilizing the strong coordination capacity of fluorine ions and thorium ions4Precipitating and separating from uranium-containing leaching solution; said H2O2The oxidability is improved, and the oxidative leaching of low-valence uranium is promoted; at this time, the H2SO4Not only can provide an acidic environment, is used for dissolving the oxidizing gangue components such as K, Mg, Al, U, Th and the like in the ferrocolumbium concentrate, but also can be mixed with H2O2The synergistic effect improves the oxidability and promotes the oxidative leaching of the low-valence uranium.
In the invention, when the acid mixing system is HF-HCl, the Cl is-The concentration of (b) is preferably 10 to 50g/L, more preferably 20 to 40g/L, and still more preferably 30 to 35 g/L. In the invention, when the acid mixing system is HF-HCl, the strong corrosivity of the acid fluoride ions is utilized to slowly corrode and dissolve the ferrocolumbium concentrate particlesThe surface and the gap interface increase the specific surface area of ore particles and activate the state of the particle interface, thereby promoting the leaching of uranium; meanwhile, thorium in the ore is converted into stable ThF by utilizing the strong coordination capacity of fluorine ions and thorium ions4Precipitating and separating from uranium-containing leaching solution; the HCl provides an acidic environment for dissolving the oxidized gangue components in the ferrocolumbium concentrate, such as mineral phases containing K, Mg, Al, etc.
In the invention, the mixed acid system is HF-HCl-H2O2When being in the presence of Cl-The concentration of (b) is preferably 10-50 g/L, more preferably 20-40 g/L, and even more preferably 30-35 g/L; h used for preparing leachate2O2Meter, said H2O2The concentration of (b) is preferably 10 to 30g/L, more preferably 15 to 30g/L, and still more preferably 20 to 25 g/L. According to the invention, the strong corrosivity of the acidic fluoride ions is utilized to slowly corrode and dissolve the surfaces of the niobium-iron ore concentrate particles and the gap interfaces, so that the specific surface area of the ore particles is increased, and the state of the particle interfaces is activated, thereby promoting the leaching of uranium; meanwhile, thorium in the ore is converted into stable ThF by utilizing the strong coordination capacity of fluorine ions and thorium ions4Precipitating and separating from uranium-containing leaching solution; said H2O2The oxidability is improved, and the oxidative leaching of low-valence uranium is promoted; the HCl can provide an acidic environment for dissolving the oxidative gangue components in the ferrocolumbium concentrate, such as mineral phases containing K, Mg, Al and the like, and can be mixed with H2O2The synergistic effect improves the oxidability of the leaching solution and promotes the oxidative leaching of the low-valence uranium.
The source of each raw material for preparing the leaching agent containing acidic fluoride ions is not particularly limited in the present invention, and commercially available products known to those skilled in the art may be used. The preparation method of the leaching agent containing the acidic fluoride ions is not specially limited, and all the components are uniformly mixed and can reach the required concentration content.
In the present invention, the volume ratio of the mass of the ferrocolumbium concentrate to the leaching agent containing acidic fluoride ions is preferably 1 g: (5-30) mL, more preferably 1 g: (10-30) mL, more preferably 1 g: (15-25) mL. In the invention, when the ratio of the mass of the ferrocolumbium concentrate to the volume of the leaching agent containing acidic fluoride ions is in the above range, the leaching agent containing acidic fluoride ions can better play a role, and the uranium in the ferrocolumbium concentrate is removed.
In the invention, the number of the leaching stages is preferably more than two stages, more preferably 2 to 5 stages, and even more preferably 3 stages. In the invention, the temperature of the multistage leaching is preferably 60-95 ℃ independently, and more preferably 70-95 ℃; the leaching time of each stage in the multi-stage leaching is preferably 3-8 hours independently, and more preferably 4-6 hours. In the present invention, the multistage leaching is preferably performed under agitation. In the present invention, the stirring rate is preferably 200 to 500rpm, and more preferably 300 to 500 rpm. In the invention, compared with one-stage leaching, the leaching agent containing acidic fluoride ions can be more fully utilized by adopting multi-stage leaching, and the leaching rate is further improved.
The present invention is not particularly limited to the specific operation of the multistage leaching, and the multistage leaching operation known to those skilled in the art may be used. In the present invention, the operation of each stage of the multi-stage leaching is preferably the same, including the concentration of the acidic fluoride leaching system, the volume ratio of the mass of ferrocolumbium concentrate to the acid fluoride-containing leaching agent, the temperature and time of leaching, and the agitation rate.
After the multi-stage leaching of the uranium is completed, the invention preferably sequentially filters and washes the products of the multi-stage leaching to obtain uranium-containing leachate and leaching slag. The operation of the filtration and washing in the present invention is not particularly limited, and filtration and washing well known in the art may be used. In the present invention, the washing liquid used for the washing preferably includes H2SO4And an aqueous solution of at least one of HF. In the invention, the mass fraction of the washing liquid is preferably 1-2%.
After leaching slag is obtained, the thorium in the leaching slag is leached in multiple stages by using ammonium salt solution, and thorium-containing leaching solution and ferrocolumbium concentrate subjected to uranium and thorium removal are obtained. In the invention, the leaching slag is leached by adopting ammonium salt solution, and the ThF is leached by utilizing ammonium ions4Strong dissolution ofAnd selective leaching and removing of thorium is realized.
In the present invention, the ammonium salt in the ammonium salt solution preferably includes at least one of ammonium carbonate, ammonium bicarbonate, ammonium sulfate, ammonium bisulfate, ammonium sulfite, and ammonium chloride. In the present invention, NH in the ammonium salt solution4 +The concentration of (B) is preferably 0.5 to 3.0mol/L, more preferably 1 to 2.5mol/L, and still more preferably 1.5 to 2.0 mol/L. The solvent of the ammonium salt solution is not specially limited, and NH in the ammonium salt solution can be ensured4 +The concentration requirement is met. In the present invention, the solvent is preferably water. In the present invention, the ammonium salt solution can dissolve ThF in the leached slag4And selective leaching and removing of thorium in the concentrate is realized. In the present invention, NH in the ammonium salt solution4 +When the concentration of (3) is within the above range, ThF in the leaching residue can be dissolved more favorably4And the thorium removal rate is improved.
The source of the ammonium salt is not particularly limited in the present invention, and a commercially available product known to those skilled in the art may be used. The preparation method of the ammonium salt solution is not specially limited, and all the components are uniformly mixed and can reach the required concentration content.
In the invention, the volume ratio of the mass of the leaching residue to the ammonium salt solution is preferably 1 g: (5-30) mL, more preferably 1 g: (10-30) mL, more preferably 1 g: (15-25) mL. In the invention, when the mass ratio of the leaching slag to the volume of the ammonium salt solution is in the range, the ammonium salt solution can better play a role in removing thorium in the ferrocolumbium concentrate.
In the invention, the number of the leaching stages is preferably more than two stages, more preferably 2 to 5 stages, and even more preferably 3 stages. In the invention, the temperature of the multistage leaching is preferably 60-95 ℃, and more preferably 70-90 ℃; the leaching time of each stage in the multi-stage leaching is preferably 20-120 min, more preferably 30-100 min, and even more preferably 50-60 min. In the present invention, the multistage leaching is preferably performed under agitation. In the present invention, the stirring rate is preferably 200 to 500rpm, and more preferably 300 to 500 rpm. In the invention, compared with one-stage leaching, the multi-stage leaching can more fully utilize the ammonium salt solution, and further improves the leaching rate.
The present invention is not particularly limited to the specific operation of the multistage leaching, and the multistage leaching operation known to those skilled in the art may be used. In the present invention, the operation of each stage of the multi-stage leaching is preferably the same, including the concentration of the ammonium salt solution, the mass of the leaching residue to volume ratio of the ammonium salt solution, the temperature and time of leaching, and the agitation rate.
After the multistage leaching of thorium is completed, the multistage leaching product is preferably sequentially filtered, washed and dried to obtain thorium-containing leaching solution and uranium and thorium-removed ferroniobium concentrate. The operation of filtering, washing and drying is not particularly limited in the present invention, and may be performed by a filtering, washing and drying operation well known in the art. In the present invention, the washing liquid used for the washing is preferably an aqueous solution containing at least one of ammonium carbonate and ammonium sulfate. In the invention, the mass fraction of the washing liquid is preferably 1-2%.
According to the method, the niobium-iron ore concentrate is leached by using the leaching agent containing acidic fluoride ions, the surfaces and the gap interfaces of niobium-iron ore concentrate particles are slowly corroded and dissolved by using the strong corrosivity of an acidic fluoride ion solution, the specific surface area of the ore particles is increased, and the state of the particle interfaces is activated, so that the leaching of uranium is promoted; meanwhile, thorium in the ore is converted into stable ThF by utilizing the strong coordination capacity of fluorine ions and thorium ions4Precipitating and separating from uranium-containing leaching solution; leaching the leaching slag by using ammonium salt solution, and leaching ThF by using ammonium ions4The strong solubility of the thorium realizes the selective leaching removal of the thorium; the method provided by the invention is simple in process and convenient to operate, and can selectively remove and separate the radioactive elements uranium and thorium in the ferrocolumbium concentrate by utilizing a hydrometallurgy process, thereby laying a foundation for further realizing the nonradioactive smelting of ferrocolumbium and the recovery of uranium and thorium.
The technical solution of the present invention will be clearly and completely described below with reference to the embodiments of the present invention. It is to be understood that the described embodiments are merely exemplary of the invention, and not restrictive of the full scope of the invention. All other embodiments, which can be derived by a person skilled in the art from the embodiments given herein without making any creative effort, shall fall within the protection scope of the present invention.
Example 1
The ferrocolumbium concentrate used in this example was purchased from Brazilian niobium mine, which contained 0.27% uranium and 1.03% thorium, and had a particle size of 10 to 80 μm (this ferrocolumbium concentrate was used in examples 1 to 21).
Step (1): leaching uranium in the ferrocolumbium concentrate by using a leaching agent containing acidic fluoride ions for two-stage leaching, filtering, and washing by using an HF solution with the mass fraction of 1% to obtain a uranium-containing leaching solution and leaching residues; wherein the volume ratio of the mass of the niobium-iron ore concentrate to the leaching agent containing acidic fluoride ions is 1 g: 10mL, 30g/LHF, 60g/LH2SO4And 15g/LH2O2Preparing a leaching agent containing acidic fluoride ions, wherein the concentration of the fluoride ions in the leaching agent containing the acidic fluoride ions is 30g/L, SO4 2-Has a concentration of 60g/L as H2O2Total mass meter, H2O2The concentration of (A) is 15g/L, and the rest is water; the temperature of each leaching stage is 60 ℃, the stirring speed in the leaching process of each stage is 300rpm, and the leaching time of each stage is 4 hours;
(2) step (2): carrying out three-stage leaching on thorium in the leaching slag obtained in the step (1) by using a 1mol/L ammonium carbonate aqueous solution, filtering, washing by using an ammonium carbonate solution with the mass fraction of 1%, and drying to obtain a thorium-containing leaching solution and niobium-iron concentrate subjected to uranium and thorium removal; wherein the volume ratio of the mass of the leaching residue to the ammonium salt solution is 1 g: 10mL, the temperature of each leaching stage is 90 ℃, the stirring speed in the leaching process of each stage is 300rpm, and the leaching time of each stage is 60 min.
Respectively measuring the uranium concentration and the thorium concentration of the uranium-containing leachate and the thorium-containing leachate obtained in the steps (1) and (2);
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 95.19 percent, and the leaching rate of thorium is 0.12 percent; the leaching rate of thorium in the thorium-containing leachate is 98.18%, and the leaching rate of uranium is 0.02%, wherein fig. 1 is a variation curve of the leaching rates of uranium and thorium with leaching time in the two-stage leaching process of uranium in example 1; FIG. 2 is a plot of the uranium and thorium leaching rates as a function of leaching time during the three-stage leaching of the thorium of example 1. According to the figure 1, the leaching rate of uranium can be greatly improved by two-stage leaching, and the leaching rate of thorium is not greatly influenced; according to the figure 2, the leaching rate of thorium can be greatly improved by three-stage leaching, and the leaching rate of uranium is not greatly influenced; it can be seen from a combination of fig. 1 and 2 that the present invention achieves the separation of uranium and thorium.
Calculating the separation rate:
the separation rate is defined as the average value of the leaching rate of uranium in the uranium removal process and the leaching rate of thorium in the thorium removal process:
Figure BDA0002602539280000091
calculated separation rate of uranium and thorium is 96.7%.
Example 2
The uranium leaching was carried out according to the method of example 1, wherein the concentration of HF was controlled to 10g/L, the concentration of fluoride ion in the acid fluoride-containing leaching agent was controlled to 10g/L, and the rest of the conditions were unchanged;
leaching of thorium was carried out according to the method of example 1, in which the concentration of ammonium carbonate was controlled to 0.5mol/L (NH)4 +The concentration of (1) mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 94.24 percent, and the leaching rate of thorium is 0.08 percent; the leaching rate of thorium in the thorium-containing leaching solution is 97.86 percent, and the leaching rate of uranium is 0.05 percent.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was found to be 96.1%.
Example 3
The uranium leaching was carried out according to the method of example 1, wherein the concentration of HF was controlled to 50g/L, the concentration of fluoride ion in the acid fluoride-containing leaching agent was controlled to 50g/L, and the rest of the conditions were unchanged;
leaching of thorium was carried out as in example 1, with the concentration of ammonium carbonate controlled to be 1.5mol/L (NH)4 +The concentration of (b) is 3.0mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 98.36 percent, and the leaching rate of thorium is 0.10 percent; the leaching rate of thorium in the thorium-containing leaching solution is 99.22%, and the leaching rate of uranium is 0.03%.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was 98.8%.
Example 4
The leaching of uranium was carried out according to the method of example 1, with H being controlled2SO4Is 30g/L, contains SO in the leaching agent of acidic fluorinion4 2-The concentration of (A) is 30g/L, and the rest conditions are unchanged;
leaching of thorium was carried out according to the method of example 1, in which the ammonium salt used was ammonium bicarbonate, the concentration of which was controlled to be 2mol/L (NH)4 +The concentration of (b) is 2mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 93.78 percent, and the leaching rate of thorium is 0.13 percent; the leaching rate of thorium in the thorium-containing leaching solution is 98.12%, and the leaching rate of uranium is 0.02%.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was found to be 96.0%.
Example 5
The leaching of uranium was carried out according to the method of example 1, with H being controlled2SO4Is 100g/L, and contains SO in the leaching agent of acidic fluorinion4 2-The concentration of (A) is 100g/L, and the rest conditions are unchanged;
thorium leaching was carried out according to the method of example 1, wherein the ammonium salt used was ammonium sulphate, the concentration of which was controlled to be 1.5mol/L (NH)4 +The concentration of (b) is 3mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 96.54 percent, and the leaching rate of thorium is 0.08 percent; the leaching rate of thorium in the thorium-containing leaching solution is 99.37 percent, and the leaching rate of uranium is 0.02 percent.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was 98.0%.
Example 6
The leaching of uranium was carried out according to the method of example 1, with H being controlled2O2Has a concentration of 10g/L and contains acidic fluoride ions in the presence of H2O2Total mass meter, H2O2The concentration of (A) is 10g/L, and the rest conditions are unchanged;
leaching of thorium was carried out according to the method of example 1, in which the ammonium salt used was ammonium bicarbonate, the concentration of which was controlled to be 0.5mol/L (NH)4 +The concentration of (b) is 0.5mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 94.98 percent, and the leaching rate of thorium is 0.05 percent; the leaching rate of thorium in the thorium-containing leaching solution is 97.52%, and the leaching rate of uranium is 0.01%.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was found to be 96.3%.
Example 7
The leaching of uranium was carried out according to the method of example 1, with H being controlled2O2Has a concentration of 30g/L and contains acidic fluoride ions in the presence of H2O2Total mass meter, H2O2The concentration of (A) is 10g/L, and the rest conditions are unchanged;
leaching of thorium was carried out according to the method of example 1, in which the ammonium salt used was ammonium bicarbonate, the concentration of which was controlled to be 3.0mol/L (NH)4 +The concentration of (b) is 3.0mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 95.88 percent, and the leaching rate of thorium is 0.05 percent; the leaching rate of thorium in the thorium-containing leaching solution is 99.12 percent, and the leaching rate of uranium is 0.02 percent.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was 97.5%.
Example 8
Pressing to realThe uranium leaching was carried out according to the method of example 1, using 30g/LHF and 60g/LH2SO4Preparing a leaching agent containing acidic fluoride ions, wherein the concentration of the fluoride ions in the leaching agent containing the acidic fluoride ions is 30g/L, SO4 2-The concentration of (A) is 60g/L, the rest is water, and the rest conditions are unchanged;
leaching of thorium was carried out according to the method of example 1, wherein the ammonium salt used was ammonium bisulfate, the concentration thereof was controlled to be 0.5mol/L (NH)4 +The concentration of (b) is 0.5mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 93.78 percent, and the leaching rate of thorium is 0.03 percent; the leaching rate of thorium in the thorium-containing leaching solution is 95.08 percent, and the leaching rate of uranium is 0.04 percent.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and found to be 94.4%.
Example 9
The uranium leaching was carried out according to the method of example 8, in which the concentration of HF was controlled to 10g/L, the concentration of fluoride ion in the acid fluoride-containing leaching agent was controlled to 10g/L, and the rest of the conditions were unchanged;
leaching of thorium was carried out according to the method of example 1, wherein the ammonium salt used was ammonium bisulfate, the concentration thereof was controlled to be 3.0mol/L (NH)4 +The concentration of (b) is 3.0mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 91.64 percent, and the leaching rate of thorium is 0.02 percent; the leaching rate of thorium in the thorium-containing leaching solution is 98.43 percent, and the leaching rate of uranium is 0.04 percent.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was found to be 95.0%.
Example 10
The leaching of uranium was carried out according to the method of example 8, with H being controlled2SO430g/L of SO in leaching agent containing acidic fluoride ions4 2-The concentration of (A) is 30g/L, and the rest conditions are unchanged;
leaching of thorium was carried out as in example 1, wherein the ammonium salt used was sulfurous acidAmmonium, controlled at a concentration of 0.25mol/L (NH)4 +The concentration of (b) is 0.5mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching liquid is 95.44 percent, and the leaching rate of thorium is 0.05 percent; the leaching rate of thorium in the thorium-containing leaching solution is 93.16%, and the leaching rate of uranium is 0.01%.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was found to be 94.3%.
Example 11
The leaching of uranium was carried out according to the method of example 8, with H being controlled2SO4SO in leaching agent with concentration of 100g/L and containing acidic fluoride ions4 2-The concentration of (A) is 100g/L, and the rest conditions are unchanged;
leaching of thorium was carried out according to the method of example 1, wherein the ammonium salt used was ammonium sulfite, the concentration of which was controlled to be 1.5mol/L (NH)4 +The concentration of (b) is 3.0mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 95.47 percent, and the leaching rate of thorium is 0.04 percent; the leaching rate of thorium in the thorium-containing leaching solution is 97.26%, and the leaching rate of uranium is 0.01%.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was found to be 96.4%.
Example 12
The leaching of uranium was carried out according to the method of example 1, with 30g/L HF and 15g/L H2O2Preparing a leaching agent containing acidic fluoride ions, wherein the concentration of the fluoride ions in the leaching agent containing the acidic fluoride ions is 30g/L, and H is used2O2Total mass meter, H2O2The concentration of (A) is 15g/L, the rest is water, and the rest conditions are unchanged;
leaching of thorium was carried out according to the method of example 1, wherein the ammonium salt used was ammonium chloride, the concentration of which was controlled to be 0.5mol/L (NH)4 +The concentration of (b) is 0.5mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching liquid is 95.44 percent, and the leaching rate of thorium is 0.11 percent; the leaching rate of thorium in the thorium-containing leaching solution is 92.36 percent, and the leaching rate of uranium is 0.08 percent.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was 93.9%.
Example 13
The leaching of uranium was carried out according to the method of example 12, wherein H2O2Is controlled to be 10g/L, and H is contained in the leaching agent containing acidic fluoride ions2O2Total mass meter, H2O2The concentration of (A) is 10g/L, and the rest conditions are unchanged;
leaching of thorium was carried out according to the method of example 1, wherein the ammonium salt used was ammonium chloride, the concentration of which was controlled to be 3mol/L (NH)4 +The concentration of (b) is 3.0mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 92.18 percent, and the leaching rate of thorium is 0.14 percent; the leaching rate of thorium in the thorium-containing leaching solution is 96.76%, and the leaching rate of uranium is 0.05%.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was found to be 94.5%.
Example 14
The leaching of uranium was carried out according to the method of example 12, wherein H2O2Is controlled to be 30g/L, and H is contained in the leaching agent containing acidic fluoride ions2O2Total mass meter, H2O2The concentration of (A) is 30g/L, and the rest conditions are unchanged;
leaching of thorium was carried out according to the method of example 1, in which the ammonium salt used was ammonium chloride, the concentration of which was controlled to be 1mol/L (NH)4 +The concentration of (1.0 mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 95.72 percent, and the leaching rate of thorium is 0.14 percent; the leaching rate of thorium in the thorium-containing leaching solution is 92.30 percent, and the leaching rate of uranium is 0.02 percent.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and found to be 94.0%.
Example 15
The uranium leaching was carried out according to the method of example 1, using 30g/LHF, 15g/LH2O2And 30g/L HCl to prepare a leaching agent containing acid fluoride ions, wherein the concentration of the fluoride ions in the leaching agent containing the acid fluoride ions is 30g/L, and H is used2O2Total mass meter, H2O2Has a concentration of 15g/L, Cl-The concentration of (A) is 60g/L, the rest is water, and the rest conditions are unchanged;
thorium leaching was carried out according to the method of example 1, wherein the ammonium salt used was ammonium sulphate, the concentration of which was controlled to be 1mol/L (NH)4 +The concentration of (b) is 2.0mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 96.70 percent, and the leaching rate of thorium is 0.02 percent; the leaching rate of thorium in the thorium-containing leaching solution is 96.26%, and the leaching rate of uranium is 0.03%.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was found to be 96.5%.
Example 16
The leaching of uranium was carried out according to the method of example 1, using 30g/L HF and 30g/L HCl to formulate a leaching agent containing acidic fluoride ions, wherein the concentration of fluoride ions in the leaching agent containing acidic fluoride ions was 30g/L, and Cl was present-The concentration of (A) is 30g/L, and the rest conditions are unchanged;
thorium leaching was carried out according to the method of example 1, wherein the ammonium salt used was ammonium sulphate, the concentration of which was controlled to be 1mol/L (NH)4 +The concentration of (b) is 2.0mol/L), and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 96.89%, and the leaching rate of thorium is 0.02%; the leaching rate of thorium in the thorium-containing leaching solution is 96.58 percent, and the leaching rate of uranium is 0.02 percent.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was found to be 96.7%.
Example 17
The leaching of uranium was carried out according to the method of example 1, with the ratio of the mass of ferrocolumbium concentrate to the volume of leaching agent containing acidic fluoride being 1: 30mL, and the rest conditions are unchanged;
thorium was leached as in example 1, the ratio of the mass of the leaching residue to the volume of the ammonium salt solution being 1 g: 30mL, and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 97.22 percent, and the leaching rate of thorium is 0.05 percent; the leaching rate of thorium in the thorium-containing leaching solution is 98.79 percent, and the leaching rate of uranium is 0.11 percent.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was 98.0%.
Example 18
Leaching uranium according to the method of the embodiment 1, controlling the temperature of each leaching section to be 95 ℃, and keeping the rest conditions unchanged;
leaching thorium according to the method of the embodiment 1, controlling the temperature of each stage of leaching to be 60 ℃, and keeping the rest conditions unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 99.13 percent, and the leaching rate of thorium is 0.18 percent; the leaching rate of thorium in the thorium-containing leaching solution is 90.44%, and the leaching rate of uranium is 0.02%.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and found to be 94.8%.
Example 19
Leaching uranium according to the method of the embodiment 1, controlling the stirring speed in each leaching process to be 200rpm, and keeping the rest conditions unchanged;
leaching thorium according to the method of the embodiment 1, controlling the stirring speed in each leaching process to be 200rpm, and keeping the rest conditions unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 95.21 percent, and the leaching rate of thorium is 0.08 percent; the leaching rate of thorium in the thorium-containing leaching solution is 96.86 percent, and the leaching rate of uranium is 0.03 percent.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was found to be 96.0%.
Example 20
Leaching uranium according to the method of the embodiment 1, controlling the leaching time of each section to be 3h, and keeping the rest conditions unchanged;
leaching thorium according to the method of the embodiment 1, controlling the leaching time of each stage to be 20min, and keeping the rest conditions unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 93.54 percent, and the leaching rate of thorium is 0.03 percent; the leaching rate of thorium in the thorium-containing leaching solution is 96.38 percent, and the leaching rate of uranium is 0.02 percent.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was found to be 95.0%.
Example 21
Leaching of uranium was carried out according to the method of example 1;
leaching thorium according to the method of the embodiment 1, wherein the number of leaching stages is 5 stages, and the rest conditions are unchanged;
the result shows that the leaching rate of uranium in the uranium-containing leaching solution is 95.22 percent, and the leaching rate of thorium is 0.14 percent; the leaching rate of thorium in the thorium-containing leaching solution is 99.96 percent, and the leaching rate of uranium is 0.09 percent.
Calculating the separation rate: the uranium and thorium separation rate was calculated according to the method of example 1, and the separation rate was 97.6%.
From the above examples, it can be seen that the method provided by the invention enables the removal rates of uranium and thorium in the ferrocolumbium concentrate to be higher than 90% and the separation rates to be higher than 90%.
The foregoing is only a preferred embodiment of the present invention, and it should be noted that, for those skilled in the art, various modifications and decorations can be made without departing from the principle of the present invention, and these modifications and decorations should also be regarded as the protection scope of the present invention.

Claims (4)

1. A method for removing and separating uranium and thorium from niobium-iron concentrate comprises the following steps:
(1) leaching uranium in the ferrocolumbium concentrate in multiple stages by using a leaching agent containing acidic fluoride ions to obtain a uranium-containing leachate and leaching slag;
(2) carrying out multistage leaching on thorium in the leaching slag obtained in the step (1) by using an ammonium salt solution to obtain a thorium-containing leaching solution and ferrocolumbium concentrate subjected to uranium and thorium removal;
the leaching agent containing acidic fluoride ions in the step (1) is a mixed acid system which is HF-H2SO4、HF-H2O2、HF-H2SO4-H2O2HF-HCl and HF-HCl-H2O2One of (1);
the concentration of the fluorine ions in the leaching agent containing the acidic fluorine ions in the step (1) is 10-50 g/L;
the temperature of the multistage leaching in the step (1) and the temperature of the multistage leaching in the step (2) are 60-95 ℃ independently;
the number of the leaching stages in the step (1) and the leaching stages in the step (2) is more than two independently;
the leaching time of each stage in the multi-stage leaching in the step (1) is independently 3-8 h, and the leaching time of each stage in the multi-stage leaching in the step (2) is independently 20-120 min;
the volume ratio of the mass of the ferrocolumbium concentrate in the step (1) to the volume of the leaching agent containing acidic fluoride ions is 1 g: (5-30) mL.
2. The method according to claim 1, wherein the ammonium salt in the ammonium salt solution in the step (2) comprises at least one of ammonium carbonate, ammonium bicarbonate, ammonium sulfate, ammonium bisulfate, ammonium sulfite, and ammonium chloride.
3. The method of claim 1, wherein in step (2), NH is present in the ammonium salt solution4 +The concentration of (b) is 0.5 to 3.0 mol/L.
4. The method according to claim 1, characterized in that the ratio of the mass of the leached residue to the volume of the ammonium salt solution in step (2) is 1 g: (5-30) mL.
CN202010729164.3A 2020-07-27 2020-07-27 Method for removing and separating uranium and thorium from ferrocolumbium concentrate Active CN111826521B (en)

Priority Applications (1)

Application Number Priority Date Filing Date Title
CN202010729164.3A CN111826521B (en) 2020-07-27 2020-07-27 Method for removing and separating uranium and thorium from ferrocolumbium concentrate

Applications Claiming Priority (1)

Application Number Priority Date Filing Date Title
CN202010729164.3A CN111826521B (en) 2020-07-27 2020-07-27 Method for removing and separating uranium and thorium from ferrocolumbium concentrate

Publications (2)

Publication Number Publication Date
CN111826521A CN111826521A (en) 2020-10-27
CN111826521B true CN111826521B (en) 2022-02-01

Family

ID=72926374

Family Applications (1)

Application Number Title Priority Date Filing Date
CN202010729164.3A Active CN111826521B (en) 2020-07-27 2020-07-27 Method for removing and separating uranium and thorium from ferrocolumbium concentrate

Country Status (1)

Country Link
CN (1) CN111826521B (en)

Families Citing this family (3)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
CN113628778B (en) * 2021-06-17 2023-12-08 福建工程学院 Method for transferring and adsorbing radioactive elements in tailings in high-cold and high-altitude areas
CN115798771B (en) * 2022-11-18 2024-02-23 中核第四研究设计工程有限公司 Method for stabilizing acidic associated radioactive waste residues and stabilizing agent used in method
CN117695844A (en) * 2024-02-05 2024-03-15 北京先通国际医药科技股份有限公司 Method for extracting decay daughter of thorium element

Citations (7)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
DE3211286A1 (en) * 1981-04-02 1982-10-21 Hermann C. Starck Berlin, 1000 Berlin Process for recovering mineral acid compounds free from radiation from mineral-acid ores containing, in complex form, uranium, thorium, titanium and rare earths
US5023059A (en) * 1988-11-02 1991-06-11 Bielecki Edwin J Recovery of metal values and hydrofluoric acid from tantalum and columbium waste sludge
CN1378604A (en) * 2000-09-12 2002-11-06 三井金属鋐业株式会社 Method for producing metal or metal compound comprising process of treating with fluorine and adjusted raw material used therein
CN102312094A (en) * 2010-07-07 2012-01-11 核工业北京地质研究院 Method for extracting uranium from uranium-contained niobium-tantalum leached tailings
CN104597174A (en) * 2013-11-04 2015-05-06 青岛齐力铸钢有限公司 Isotope dilution mass spectrometry method for determining content of uranium in uranium niobium alloy
CN104775026A (en) * 2015-03-31 2015-07-15 江西洁球环保科技有限公司 Method for extracting high-purity uranium, thorium and mixed rare earths from excellent molten slag
CN106676291A (en) * 2016-12-28 2017-05-17 核工业北京化工冶金研究院 Method for comprehensively recovering uranium, niobium and tantalum from ore

Patent Citations (7)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
DE3211286A1 (en) * 1981-04-02 1982-10-21 Hermann C. Starck Berlin, 1000 Berlin Process for recovering mineral acid compounds free from radiation from mineral-acid ores containing, in complex form, uranium, thorium, titanium and rare earths
US5023059A (en) * 1988-11-02 1991-06-11 Bielecki Edwin J Recovery of metal values and hydrofluoric acid from tantalum and columbium waste sludge
CN1378604A (en) * 2000-09-12 2002-11-06 三井金属鋐业株式会社 Method for producing metal or metal compound comprising process of treating with fluorine and adjusted raw material used therein
CN102312094A (en) * 2010-07-07 2012-01-11 核工业北京地质研究院 Method for extracting uranium from uranium-contained niobium-tantalum leached tailings
CN104597174A (en) * 2013-11-04 2015-05-06 青岛齐力铸钢有限公司 Isotope dilution mass spectrometry method for determining content of uranium in uranium niobium alloy
CN104775026A (en) * 2015-03-31 2015-07-15 江西洁球环保科技有限公司 Method for extracting high-purity uranium, thorium and mixed rare earths from excellent molten slag
CN106676291A (en) * 2016-12-28 2017-05-17 核工业北京化工冶金研究院 Method for comprehensively recovering uranium, niobium and tantalum from ore

Also Published As

Publication number Publication date
CN111826521A (en) 2020-10-27

Similar Documents

Publication Publication Date Title
CN111826521B (en) Method for removing and separating uranium and thorium from ferrocolumbium concentrate
CN102899485B (en) Resin in pulp process is from containing the method extracting scandium scandium material
CN110055423B (en) Method for enriching platinum group metals and rare earth in spent automobile exhaust purification catalyst
CN104962762A (en) Processing method of fine bastnaesite
CN114107668A (en) Method for mechanically activating and strengthening leaching rare earth in waste polishing solution
WO2019099859A1 (en) Selective recovery of rare earth metals from an acidic slurry or acidic solution
Kursun et al. Solubility of Eskisehir thorium/rare earth ores in sulphuric and nitric acids
CN108251656A (en) The method for extracting gold, platinum and palladium in electronic waste copper anode mud
CN115852177A (en) Method for recycling scandium from fused salt chlorination dust collection slag
CN106216364A (en) A kind of SCR waste catalyst recovery method and system
CN115491527A (en) Method for pretreating uranium-containing waste residues and method for recycling uranium
RU2070596C1 (en) Method of scandium concentrates production
CN115974128B (en) Method for recycling scandium from fused salt chlorination comprehensive utilization byproducts by utilizing titanium white waste acid
CN104975192A (en) Method for extracting scandium from scandium-containing material
CN115786744A (en) Method for extracting scandium from titanium white waste acid and fused salt chlorination dust collection slag in synergy mode
CN113151695B (en) Method for decomposing low-grade tantalum-niobium resource and extracting tantalum-niobium by acid method
CN114164344B (en) Method for separating and recovering zinc and cobalt from cobalt-containing zinc slag
CN115852179A (en) Method for extracting scandium from fused salt chlorination dust collecting slag
CN1188155A (en) Method for extracting mixed rare earth oxide by salmiac roasting black weathering slime
CN115852178A (en) Method for extracting scandium from titanium white waste acid and molten salt chlorination waste salt in synergy mode
CN115896463A (en) Method for extracting scandium from titanium white waste acid and titanium extraction tailings in cooperation mode
CN115786742B (en) Comprehensive recycling method for uranium beryllium ore
CN115874070A (en) Method for extracting scandium from molten salt chlorination waste salt
CN110629025A (en) Method for efficiently leaching cobalt and nickel from slag after tungsten extraction
CN117737428A (en) Method for recycling scandium from scandium-containing solid waste generated by extracting titanium by chlorination process

Legal Events

Date Code Title Description
PB01 Publication
PB01 Publication
SE01 Entry into force of request for substantive examination
SE01 Entry into force of request for substantive examination
TA01 Transfer of patent application right
TA01 Transfer of patent application right

Effective date of registration: 20210712

Address after: 330013 No. 418, Guanglan Avenue, Jingkai District, Nanchang City, Jiangxi Province

Applicant after: EAST CHINA INSTITUTE OF TECHNOLOGY

Applicant after: GUANGDONG INSTITUTE OF MINERAL APPLICATION

Address before: 330013 No. 418, Guanglan Avenue, Jingkai District, Nanchang City, Jiangxi Province

Applicant before: EAST CHINA INSTITUTE OF TECHNOLOGY

GR01 Patent grant
GR01 Patent grant