Method for receiving actinium-227 and thorium-228 from treated by neutrons in reactor radium-226

FIELD: metallurgy.

SUBSTANCE: invention concerns manufacturing of radionuclides for industry, science, nuclear medicine, especially radioimmunotherapy. Particularly it concerns method of receiving actinium -227 and thorium -228 from treated by neutrons in reactor radium-226. Method includes irradiation of target containing of metallic capsule in which there is located reaction vessel, containing radium-226 in the form of compound. Then it is implemented unsealing of target's metallic capsule, dissolving of received radium. From solution it is separated by means of precipitation, and then it is implemented regeneration, preparation to new irradiation and extraction of actinium-227 and thorium-228 from solution. At that irradiation, dissolving, radium separation, its regeneration and preparation to new irradiation are implemented in the form of its united chemical form - radium bromide, in the same reaction vessel made of platinum. Method provides reusing of the same platinum vessel for receiving of actinium-227 and thorium-228 from one portion of radium by recycling of irradiation and extraction in the same vessel. Separation of metallic capsule by means of dissolving provides saving of mechanical integrity of platinum reaction vessel for each new irradiation cycle and extraction.

EFFECT: increasing of radiationally-environmental safety of process, excluding operations of increased radiation hazard.

2 cl, 2 ex


The invention relates to the production of radionuclides for industry, science, medicine, and especially of nuclear medicine, problems which require the use of and obtaining of radionuclides in significant quantities and a wider spectrum of radiation properties. The spread of cancer require new approaches in therapy of these diseases - of radioimmunotherapy and new radionuclides for their implementation. Such radionuclides recognized radionuclides alpha emitters, the cytotoxic effect of which is 2-3 orders of magnitude higher than the effect of beta-emitters. Of these the most promising (on the biological effects and radiotherapy efficiency) recognized Tb-149, At-211, Pb-212 (Bi-212), Ra-223, Ra-224 Ac-225, Bi-213 [MR McDevitt., Sgouros G., Firr R.D. at. al. "Eur, J. Nucl. Med." 1998. v.25. No. 9, p.1341-1351]. Three of them are products of the decay of AC-227 Ra-223) and Th-228 (Ra-224, Pb-212 (Bi-212). To obtain them in sufficient quantities at relatively low cost are more satisfied with the process of separating them from irradiated fuel by thermal neutrons in the reactor Ra-226 in response mnogoznachnogo capture

Other sources for AC-227 Th-228 rather time-consuming to produce them in sufficient quantities: Ac-227 - of the pre-received and aged drugs Pa-231 [Henrecsen G., J. Alstad, Larsen R. the Radiochim. Acta" (2001) 89, 661-666]; Ac-227 from waste and concentrates processing uranium [Bray at al. U.S. Patent 5809394, 15.09.1998. Bray at al. U.S. Patent 5885465, 23.09.1999]; Th-288 - from aged drugs U-232, obtained by neutron irradiation of Pa-231 [Deliken O. "Health Phys" (1978) v. 35 (Juli), p.21-14] or from the old salts of thorium through the preliminary allocation of Ra-228 [Sani A.R. "J. Radioanal chem.", 1970, v. 4, p.127]. These isotopes other nuclear reactions is not justified because of the small exit.

The process of allocating large amounts of AC-227 Th-228 from irradiated Ra-226 different specific features of this system: high specific power, beholden to the formation of two radioactive families (AC-227 (4n+3) and Th-228 (4n)), manifested in the radiation destruction process reagents, the radiolysis of solutions, the violation of technological regimes, the allocation of large quantities of radioactive gases (Rn-22, Tn-220, An-219) and the higher level of radiation hazard. Therefore, the urgent problem of any large-scale separation of the AC-227 Th-228 from irradiated Ra-226 is the withdrawal (separation) first Ra-226 and other radium isotopes from the process to reduce the impact of radiation on the process.

Upon receipt of small amounts of AC-227 Th-228 for the research programs of these problems do not occur.

Known gas-phase processes in particular m is re solve these problems, since it is less susceptible to radiation effects and can be implemented in a hermetically closed volume (units):

fractional sublimation halides at temperatures of 800-900°and unloading ˜10-6mm Hg [French Patent, No. 1518786, CL C22, 17.05.1968]. However, engineering and technological solutions for the implementation of such processes has not received the necessary development for use in a practical, large-scale radiochemical processes.

Closest to the claimed is Belgian process of large-scale production of AC-227 from irradiated Ra-226 isotope thermoelectric generator based on it [L.H.Baetsle, P.Dejonghe, A.C.Demild, A.De.Troyer, A.Droissart, M.Poskin "Fourth United Nation International Conference on the Peaceful uses of Atomic Energy, Geneva, 1971, A/Conf. 49/P/287. Belgium, May, 1971].

This process includes the following main steps or procedures which reflect the peculiarities of its implementation:

- Preparation of target irradiation, the target consists of a capsule, made of stainless steel, in which tight contact with the walls of the capsule is inserted aluminum matrix (Al-matrix) with holes to fill in their calcined and crushed powder of radium carbonate (RaCO3and seal it in them to a certain density (˜1,9 g/cm), the capsule is sealed (the terminology of the prototype, 2);

- About the doctrine of the target in the reactor is in the form of carbonate (RaCO 3);

- Opening of the irradiated target. Is sharp outer capsule;

- Separation of radium (from the as and Th):

- The dissolution of irradiated RaCO3directly in the capsule in the channels of the Al-matrix in dilute nitric acid (HNO3) (with the addition of acetic acid to absorb expansion). At this stage in the manufacturing process are corrosion products capsules and Al-matrix.

- Nitrate solution of radium Ra(NO3)2transferred from capsules with a vessel of tantalum for precipitation and separation of radium. Department of radium is in the form of nitrate radium by deposition of concentrated nitric acid (80% HNO3in tantalum vessel. The solution after washing the precipitate is directed to the extraction and separation of AC-227 Th-228, and the residue nitrate radium - on regeneration in carbonate form. Corrosion products capsules and Al-matrix are distributed between the precipitate Ra(NO3)2and the solution containing AC-227 Th-228.

- Regeneration of radium (preparation for new irradiation); as a form of radium irradiation is the radium carbonate and form of its branch - nitrate radium in the prototype is a series of operations on transfer of nitrate radium carbonate (RaCO3). Sediment Ra(NO3)2in tantalum vessel after separation of the filtrate is dissolved in water and transferred to another vessel for deposition RaCO .

- Deposition RaCO3by adding an excess of NH4OH to aqueous solution Ra(NO3)2while continuously passing gaseous ammonia solution CO2for sediment RaCO3. Precipitation and separation of radium in the form RaCO3is not full. The reason is the high radiation load (radiolysis, bound in the presence of all isotopes of radium Ra-226, Ra-224, Ra-223 with the products of decomposition, breaking the mode of deposition) and the presence of corrosion products capsules and Al matrix Al, Fe, Cr, Ni, Mn, etc. generated in the ammonia environment hydrolyzed and colloidal forms an exciting part of radium.

- In order to avoid losses of radium at this stage in the prototype is an additional stage fixation losses in the form of radium chromate RaCrO4that is precipitated in the filtrate after separation of the precipitate RaCO3.

- Preparation RaCO3the new irradiation:

- Sediment RaCO3in tantalum vessel is dried and made red-hot;

- Calcined and sintered sludge RaCO3crush;

Powder chopped RaCO3fill the channels of the new Al-matrix new target and use it for a new exposure.

- Separation and AC-227 Th-228 is known ion exchange method on the anion exchange resin Dawex AG 1×8 5 M HNO3.

- Clear Ac-227 is swetnam way in the form of oxalate [Scharlau, Bramasole, "Anemone", M, 1982].

The method selected as a prototype, its characteristics and structure (schema) of the technological process of selection was determined by two initial conditions: the difference of the chemical forms of radium irradiation (in the form of carbonate, RaCO3and its separation from the AC-227 Th-228 (in the form of nitrate radium, Ra(NO3)2), as well as the target device for the irradiation of radium carbonate.

The first one requires the inclusion in the technological process of recovery operations radium to prepare it in the form of carbonate to a new irradiation. This is reflected in the organization of a single phase with the appropriate infrastructure and hardware equipment. This stage due to the high radiation exposure, are obligated by the presence of all radium isotopes and their decay products, as well as the appearance of corrosion products of the target (with the dissolution of irradiated RaCO3), together with technological breaches of operations and technological losses of radium, which in turn requires the introduction of a process diagram of another self-phase - commit losses of radium in the form of chromate RaCrO4(and also with its own infrastructure).

In addition, the proposed prototype of the regeneration process according to the conditions of its implementation is accompanied by such radiation-hazardous operations as the sparging gas CO 2through ammonia solution of radium for the formation of carbonate, which is accompanied by a powerful selection of all radioactive gas (Rn-222, Tn-220, An-219) and the formation of aerosols with all isotopes of radium (Ra-226, Ra-224, Ra-223) and their degradation products. Used in the prototype device targets for irradiation carbonate necessitates the event of dissolution of irradiated radium carbonate directly in the capsule (after cutting) in the channels of the Al-matrix and the emergence in the process of corrosion products capsules and matrix. The latter are, as already mentioned, one of the reasons (along with radiolysis) of process disruption phase regeneration of radium, due to the formation of difficult-hydrolyzed and colloidal forms of the products of corrosion and loss of radium with them. The introduction of a separate commit phase, loss of radium in the form RaCrO4and at the stage of separation of the AC-227 Th-228 - refraining from the use of radiation resistant inorganic ion exchangers (phosphate Zr and Ti) due to the loss of ion exchange capacity on the corrosion products (see prototype, p.7).

The use of targets in such a device, as proposed in the prototype, accompanied inevitably by the application at the stage of preparation of the target to radiation radiation hazardous dust-raising operations: grinding sintered at proquali the years RaCO 3, transferring it from the vessel to annealing in Al-matrix and fill it with the powder channel Al-matrix and its seals. In addition, the design solution, which is implemented in the target proposed in the prototype allows only a single use of each target with the obligatory passage each time all stages of the technological process of preparation of radium for a new exposure.

The present invention is:

- reduction of mnogostadiinost process;

- increase the reliability of technological process (excluding unreliable operations for their implementation in conditions of high radiation radiation effects and can cause a process upset);

- simplifying and reducing the complexity of operations;

- improving radiation safety process: an exception or a reduction in the number of radiation-hazardous operations (aerosols, dust-raising, sparging).

To obtain such technical outcomes identified objectives, the present invention retains all of the functional stages and operations with radium, in common with the prototype, features, unlike the prototype, the use of a single chemical form of radium - radium bromide (RaBr2) - at all stages and operations with them: dissolution, separation (osaid the tion), regeneration, a new irradiation and the irradiation, and the implementation of all these transactions is proposed to carry out in the same reaction vessel made of platinum, as heat-and corrosion-resistant material, inert to activation under neutron irradiation in the reactor and before the irradiation of radium is placed to seal in a metal capsule, for example, from aluminum.

The use of a single chemical form of the bromide of radium - allows you to perform the following tasks:

- Merge stages of the separation of radium (in the form of nitrate) and regeneration (in the form of carbonate) in one chemical separation of radium (in the form of bromide) eliminates the regeneration process as a separate, independent step (stage) process; besieged RaBr2after the separation of the mother liquor, containing AC-227 Th-228 is a form that is ready for a new exposure.

- Exclusion stage of regeneration in the form, how it implemented in the prototype, avoids operations fixation losses of radium in the form RaCrO4as a separate stage of the process.

- Exclusion stage of regeneration as it is represented in the prototype allows to increase process reliability process to avoid such unreliable in conditions of high radiac the organizational impacts (radiolysis, the decomposition of the reagents) operations as precipitation of radium carbonate in an ammoniacal medium in the presence of corrosion products and education are inseparable forms of corrosion products.

To improve the radiation environment and radiation safety of the process, excluding such radiation-hazardous operations as the sparging gas CO2through ammonia solution of radium, accompanied by a strong release of radioactive gas (Rn-222, Tn-220, An-219) and the formation of aerosols with all isotopes of radium (Ra-226, Ra-224, Ra-223) and their degradation products.

- Merge stages of the separation of radium and its regeneration simplifies the process, reduces its complexity in its implementation and equipment.

Carrying out these activities, including irradiation, in the same reaction platinum vessel allows the development to the results listed previously, carry out other tasks:

- the Association of operations for the preparation of radium to irradiation (roasting, grinding) and the exposure in one of the platinum vessel allows to avoid such time-consuming stage as filling powder calcined radium narrow channels Al-matrix (diameter 5 mm) and seal them up to a certain density.

- The Association of operations of the dissolution of irradiated radium separation (sedimentation) and regeneration carried out on the prototype, three times what's the reaction vessels (dissolution - in the capsule, settling - in tantalum vessel regeneration in the third vessel) in the same platinum vessel can significantly simplify the process, to avoid operations by transferring the solution from the capsule in tantalum vessel, and into the vessel for regeneration, to thereby reduce the complexity of the process and simplify it in technological equipment and implementation.

- Use this difference allows preventing mechanical impurities and corrosion products (at the opening of the capsule and dissolve it irradiated radium) in the technological process and to avoid violating its technological mode.

To improve the radiation environment and radiation safety of the process is to avoid radiation-hazardous dust-raising operations as grinding calcined radium in the vessel for regeneration, the transfer of the powder in the Al-matrix, filling her channel and seals in them that allows you to simplify the whole process in General.

The use of platinum vessel in such a multi-purpose enables technological process of separation of AC-227 Th-228 repeatedly in the same platinum vessel with the same dose (loading) of radium in interleaving and repetition cycles of irradiation and highlight what distinguishes the claimed method from the prototype, that is for each new exposure training is required (equipment) new Al-matrix.

To achieve such a result in the proposed invention for sealing a platinum vessel is used in contrast to the prototype metal capsule, for example, of aluminum, which is placed in a platinum vessel. This allows you to open the aluminum capsule dissolution without violating the integrity of the platinum vessel, thereby providing multiple use of a single serving of radium.

The proposed method is carried out in the following sequence:

1. The irradiation of the target with radium with neutrons in the reactor (target - RaBr2in a platinum vessel in an aluminum capsule, sealed by welding).

2. The opening of the target is carried out by dissolving the aluminum capsule.

3. Department of radium.

3.1. The dissolution of irradiated RaBr2in a platinum vessel in 0.1 M HBr in minimum amount.

3.2. Deposition RaBr2concentrated HBr.

3.3. The separation of the mother liquor.

4. Preparation RaBr2the new radiation.

4.1. Drying and calcination of the precipitate RaBr2in a platinum vessel.

4.2. Sealing a platinum vessel with calcined RaBr2the new aluminum capsule.

5. Split AC-227 Th-228 is carried out by extraction of Th-228 tributyl phosphate or trioctylamine in a known mode.

6. Cleaning AC-227 is carried out by deposition in the form of oxalate. [Scharlau Bramasole, "Anemone", M, 1982].


Processing subject target (platinum vessel in an aluminum capsule containing 150 mg of Ra-226 (metal) in the form of radium bromide, 137 MCI AC-227 and 130 MCI Th-228.

Opening aluminum capsules was carried out by its dissolution in the solution composition (NaNO3+NaOH). The dissolution of irradiated RaBr2was made in a platinum vessel in 0.1 M HBr (minimal to create a saturated solution).

Department of radium is the same as for platinum deposition vessel in the form of a bromide (RaBr2) with concentrated HBr. The mother liquor containing AC-227 Th-228, is separated from the precipitate, the precipitate is washed with concentrated HBr. The combined mother liquor pariveda to convert it into nitrate form. The wet residue is dissolved in 6 M HNO3. Split AC-227, Th-228 is 40% tributyl phosphate (TBP) in benzene from aqueous 6 M HNO3. The aqueous phase containing AC-227 washed with benzene. Reextracted Th-228 from the organic phase is 0.1 M HNO3. The reextract washed with benzene. The Output Of The AC-227 - 92%, Th-228 - 81%.


Processing subject target (platinum vessel in an aluminum capsule)containing 1990 mg Ra (metal) in the form of RaBr2(radium bromide) 1880 MCI AC-227 and 1760 MCI Th-228.

The opening of the aluminum capsule is its dissolution in p is the target composition (NaOH+NaNO 3). The dissolution of irradiated radium bromide is carried out in a platinum vessel in a minimum volume of 0.1 M HBr. Department of radium is in the same platinum vessel of its precipitation in the form of the bromide with concentrated HBr. The precipitate is separated, washed with concentrated HBr, washing solutions are combined with main, mother.

The combined mother liquor containing Ac-227 Th-228, pariveda to convert it into nitrate form. The wet residue is dissolved in 3 M HNO3.

Split AC-227 Th-228 is carried out by extraction with 0.4 M solution of trioctylamine (TOA) in debutalbum the ether (the RHEED) from aqueous solution 3M HNO3. The extract containing Th-228, washed with 3M HNO3and Th-228 restrained 0.1 M HNO3, reextract washed the RHEED. The raffinate containing AC-227, washed the RHEED.

Bromide of radium (after sedimentation)in the same platinum vessel is dried, made red-hot at a temperature of ˜300°S, is closed by a stopper and sealed in a new aluminum capsule for a new exposure.

1. The method of obtaining sea anemone-227 and thorium-228 from neutron-irradiated in the reactor of radium-226 that includes the irradiation of a target consisting of a metal capsule, which placed the reaction vessel containing radium-226 in the form of a connection, opening a metal capsule of the target, the dissolution of irradiated radium, his Department is of deposition, regeneration, preparing for the new exposure and isolation of sea anemone-227 and thorium-228 from a solution, characterized in that the irradiation, dissolution, separation of radium, its regeneration and preparation for a new irradiation is carried out in the form of a single chemical forms of radium bromide, in the same reaction vessel made of platinum.

2. The method according to claim 1, characterized in that the opening of the metal capsule, in which is placed a vessel made of platinum, carry out the dissolution.


Same patents:

FIELD: nuclear medicine.

SUBSTANCE: method of realizing of neutron-catch therapy is based upon introduction of medicinal preparation into damaged organ or tissue of human body. Preparation has isotope with high cross-section of absorption of neutrons. Then damaged organ or tissue is irradiated by neutrons of nuclear reactor. Irradiation is performed with ultra-cold neutrons with energy of 10-7 eV and higher, which neutrons are released from cryogenic converter of neutrons of nuclear reactor and are delivered to damaged organ or tissue along vacuum neutron-guide, which neutron-guide has end part to be made in form of flexible catheter. Dosage loads are reduced.

EFFECT: minimized traumatism of healthy tissues of patient.

4 cl, 1 dwg, 1 tbl

FIELD: production of radioactive isotopes.

SUBSTANCE: proposed method for producing nickel-63 radioactive isotope from target within reactor includes production of nickel-62 enriched nickel target, irradiation of the latter in reactor, and enrichment of irradiated product with nickel-63, nickel-64 content in nickel-62 enriched target being not over 2%; in the course of product enrichment with nickel-63 nickel-64 isotope is extracted from irradiated product.

EFFECT: enlarged scale of production.

1 cl, 2 tbl

FIELD: radio-chemistry; methods of production of the chromatographic generator of technetium-99m from the irradiated by neutrons molybdenum-98.

SUBSTANCE: the invention is pertaining to the field of the radio-chemistry, in particular, to the methods of production of technetium-99m for medicine. Determine the specific activity of the molybdenum and the sorptive capacity of the used aluminum oxide in molybdate-ions. The mass of the molybdenum necessary for production of the preset activity of the eluate of technetium-99m determine from the ratio:ATc= 0.867·L·m ln (m)/ln(mox·Wi), where:ATc - activity of the eluate of technetium-99m, Ki; L - the specific activity of molybdenum, Ki/g; m - mass of molybdenum, g;mox - the mass of aluminum oxide in the chromatograph column, g; Wi - the sorptive capacity of the used aluminum oxide in molybdate-ions, g/g. After making of corresponding calculations the solution of molybdenum is applied on the aluminum oxide. The technical result of the invention consists in production of the generator with the required activity of technetium-99m at usage of the minimum quantity of molybdenic raw.

EFFECT: the invention ensures production of the generator with the required activity of technetium-99m at usage of the minimum quantity of molybdenic raw.

1 ex, 2 tbl, 1 dwg

The invention relates to nuclear energy, in particular the production of energy, transmutation of radioactive waste, burning weapons-grade plutonium and actinides

The invention relates to radiation technique and can be used for irradiation of internal targets

The invention relates to applied radiochemistry and relates, in particular, production facilities for extraction of the radioactive isotope carbon-14, which is widely used in the form of labeled organic compounds, as well as in the sourcesradiation

The invention relates to applied radiochemistry and relates, in particular, production to obtain a radioactive isotope of carbon14With widely used as labeled organic compounds, as well as in the sourcesradiation

The invention relates to the field of applied radiochemistry, in particular the production of radiopharmaceuticals for medicine

FIELD: chemistry.

SUBSTANCE: method involves co-sedimentation of admixture-containing solution of americium oxalate on a carrier represented by calcium oxalate, followed by obtaining nitrate americium-containing solution and americium oxalate, with its further calcination to dioxideo. Americium-containing carrier sediment is also calcinated to oxides. Nitrate solution is obtained by dissolving oxides formed during calcination in nitric acid. Americium is extracted from nitrate solution with the help of solid extragent based on diisooctylmethylphosphonate, with further re-extraction. Americium oxalate is obtained by sedimentation from condensed re-extract.

EFFECT: extended range of methods of obtaining dioxide.

3 cl, 3 ex

FIELD: metallurgy.

SUBSTANCE: invention refers to extraction and concentration of thorium out of process waste of loparit concentrates treatment - spent melt of saline sprinkler filter (SSF) of loparit concentrate chlorination process. The method includes preparation of suspension by means of discharge of spent melt of saline sprinkler filter (SSF) into water, incorporation of high molecular flocculant, of holding, filtering, separation of sediment, obtaining of chloride solution, and of treatment with steel scrap and metal magnesium. Prior to obtaining chloride solution the source suspension is heated to 60-90°C and treated with solution of sodium hydroxide to pH 1.5-2.0 and to 0.1-0.3% solution of high molecular flocculant at amount of 3-5% from the source volume of suspension; then suspension is held for 2-4 hrs. Chloride solution is received by means of filtration of spent suspension obtaining sediment of rare metals; chloride solution is then treated with steel scrap and metal magnesium; at that the solution is successively treated first with the steel scrap at amount of 3-5 mass fractions of iron per 1 fraction of iron ions (III) in chloride solution at 80-100°C for 1-3 hrs till achieving the value of pH in a pulp equal to 3.0-3.5. Then the pulp is separated from the non-reacted portion of the steel scrap and is treated with metal magnesium to pH 3.5-4.5, and further with 0.1-0.3% solution of high molecular flocculant taken at amount of 5-20% from the volume of chloride solution. Thus produced pulp is held without mixing for 1-4 hrs and filtered producing thorium containing sediment; the said sediment is washed at filter first with solution containing 1-5 g/dcm3 of sodium sulphite, then with water. Washed out sediment is repulped in solution of sodium hydroxide with concentration of 50-150g/dcm3 at a ratio of "Ж:Т"=3-5 at 60-90°C for 2-3 hrs, after what the pulp is filtered with separation of alkaline filtrate. Thorium containing sediment at the filter is washed with water, pressed at the filter and dried; the alkaline filtrate and process water are merged and mixed, then heated to 80-90°C, and treated with solution of sodium hydroxide to pH 11-13 with production of hydroxide pulp. Hydroxide pulp is filtered and then radioactive sediment is produced at the filter; it is washed out with water and transferred to a special wastes depositary, while filtrate is mixed with 10-20 volumes of shop flush water, heated to 80-90°C and again treated with solution of sodium hydroxide to pH 11-13. Obtained pulp is held and filtered thus producing sediment of rare metals and deactivated chloride solution which is discharged to drainage. Sediment of rare metals is unloaded from the filter, merged with sediment of rare metals extracted from the source suspension, dried, washed out and then transferred for preparation of charge for its further chlorination together with the loparit concentrate.

EFFECT: upgraded efficiency of thorium extraction and simultaneously solving problem of neutralisation and utilisation of process waste.

1 dwg, 1 ex

FIELD: chemistry.

SUBSTANCE: invention pertains to the technology of rare and radioactive elements; solves the problem of decomposing monazite. The method of decomposing monazite involves its treatment in molten salts at temperature ranging from 400°C to 900°C and phosphorous removal. The salts used during treatment are nitrates of alkaline metals (MeNO3), and phosphorous removal is done by separation of the clear phase of the smelt and/or lightening the phosphate of alkaline metal (Na or K) in a water solution.

EFFECT: low treatment temperature and provision for separation of phosphorous as a commercial product.

5 cl, 1 tbl, 6 ex

FIELD: metallurgy.

SUBSTANCE: said utility invention relates to hydrometallurgical methods of crude ore processing and may be used for sulphuric-acid agitation, heap, and underground leaching of uranium during uranium recovery from ores. The method involves uranium and iron leaching with sulphuric acid solution using ferric iron contained in the ore as the oxidiser; after that, uranium is recovered from the solution to prepare a solution containing ferrous iron, the ferrous iron is regenerated to ferric iron by oxidising to prepare bypass solution, and it is recirculated to the ore leaching. The uranium recovery from the solution is performed by sorption on an anion-exchange substance; after sorption, the solution containing ferric iron is acidified with sulphuric acid before the ferric iron regeneration to ferrous iron in the solution, and regeneration is performed by irradiating it with an accelerated electron flow at an absorbed dose rate of 2.3-3.5kGy/s during 1- 6 minutes.

EFFECT: increase in cost effectiveness, efficiency, and environmental safety of process.

4 cl, 3 dwg, 3 tbl, 2 ex

FIELD: production methods.

SUBSTANCE: method of monazite recycling includes the milling of the monazite, processing during the heating by substance of hydroxide of alkaline metal, generating of the salt of phosphor acid, dilution of the filter cake in the mineral acid with the following abstraction of rare earth elements (REE), thorium and uranium. Processing is done by substance of kalium hydroxide = 1:1.0-1.5 with obtaining the substance of triallyl phosphate kalium and precipitation, containing the hydroxide of thorium, uranium, REE, notopened monocyte and empty land, which is processing by azotic acid, extending the nitrate REE in the substance. It is importuned from the substance the carbonates of REE by kalium carbonate. The rest of cake is processing by the substance of kalium carbonate with translating uranium into the substance and following importuning as dihydroxide dioxuranium and the final processing of the cake by the substance of azotic acid with generating thorium into the substance by importuning of thorium by the substance of kalium carbonate. The mother water from the importuning of REE, thorium and uranium and three kalium phosphate is distained to the producing of manuring. The substance of three kalium phosphate and hydroxide kalium is vapored , and it is separated crystal three kalium phosphate, and hydroxide kalium is distained to the head of process. The rest after processing by azotic acid not opened monocyte is distained to the head of process.

EFFECT: simplifying of the process and more effective using of all components of monoyte.

6 cl, 1 ex

FIELD: technology of processing uranium-and fluorine-containing wastes.

SUBSTANCE: proposed method includes preparation of solutions from wastes, concentration of solutions by sedimentation of uranium followed by dissolving of sediments in nitric acid, extraction conversion of concentrated solutions with the use of tributyl phosphate in hydrocarbon thinner and sedimentation of ammonium polyuranates from re-extracts thus obtained. Sedimentation of uranium at stage of concentration is performed with the use of sodium hydroxide at pH= 9-10 and temperature of 60-90C. Proposed method enhances purification of uranium from fluorine due to enhanced sedimentation and filtration properties of sediments at concentration stage. Content of admixtures in triuranium octa-oxide powders obtained from re-extracts by sedimentation of ammonium polyuranates and subsequent calcination does not exceed specified norms.

EFFECT: enhanced efficiency.

1 dwg, 2 tbl, 1 ex

FIELD: processing uranium-containing products formed at extraction of uranium from solutions followed by re-extraction by means of ammonium carbonates; extraction of uranium and accompanying valid components from ores.

SUBSTANCE: proposed method includes thermal dissociation at sedimentation of uranium, entrapping of ammonia and carbon dioxide from waste gases. Thermal dissociation of uranium-containing ammonium carbonate solutions is performed at temperature of 70-85°C to pH= 6.5-5.9 at simultaneous blowing of gases by air; solutions obtained after thermal dissociation are separated from uranium-containing sediment and accompanying valid components, molybdenum for example are extracted from them.

EFFECT: enhanced efficiency of utilization of ammonia and carbon dioxide; high degree of separation of uranium and admixtures; extraction of accompanying valid components, molybdenum for example.

2 cl, 1 tbl, 2 ex

FIELD: hydraulic metallurgy.

SUBSTANCE: method comprises extracting saturated ionite from the pulp, washing it with water, desorbing uranium, washing desorbed ionite to decrease acidity, separating by wet screening into 1.0±0.2-mm size, extracting silicon from the under-screen product, and discharging it and above-screen product to the uranium sorption.

EFFECT: reduced ionite consumption.

1 cl, 1tbl

FIELD: chemistry.

SUBSTANCE: extractant has bi-phosphorus acid and additionally has tri-phosphate with relation of said components (0,5-1,25):1. Method for producing extractant includes adding to 2-ethylhexanole of chlorine oxide of phosphorus with their relation (4,5-5,1):2,0, and with parameters determined by reaching fullness of passing of reaction, after that reaction mixture is exposed until full removal of formed chlorine hydrogen, then to received mixture 1 mole of water is added, mixture is exposed to full hydrolysis. Then mixture is washed ad water layer is separated from organic remainder, containing said bi-phosphoric acid and tri-phosphate.

EFFECT: higher efficiency.

2 cl, 1 dwg, 2 tbl, 4 ex

FIELD: uranium technologies.

SUBSTANCE: method comprises sorption of uranium on low-basicity anionites, desorption of uranium, and recovery of finished product. In particular, uranium-saturated low-basicity anionite is converted into OH- form and uranium into soluble stable complex [UO2(CO3)3]-4 by treating sorbents with alkali metal and ammonium carbonate solutions.

EFFECT: achieved complete desorption of uranium and simultaneously sorbent is freed from poisons and other sorption components.

2 dwg

FIELD: metallurgy.

SUBSTANCE: invention concerns method of gallium extraction from aluminate mud. Method includes electrodialysis, in which solution for recapture containing gallium is subjected to electrodialysis for gallium concentration and acid extraction, stage of iron removal, when gallium is removed from concentrated solution, and stage of ultrafiltration, when iron-free gallium solution is neutralized. Received suspension of gallium hydroxide is subjected to ultrafiltration for suspension concentration. Then it is implemented stage of resolution, when concentrated suspension of gallium hydroxide is dissolved in alkaline solution, and electrolysis stage, when alkaline electrolytic solution of gallium received on stage of resolution is subjected to electrolysis for metallic gallium extraction.

EFFECT: providing of ability to reuse acid excluding waste acid generation commercially, increasing of gallium concentration in alkaline electrolytic solution of gallium, thereby increasing current efficiency on the electrolysis stage, and decreasing of spent solution quantity.

5 cl, 1 dwg, 2 tbl, 1 ex