RU2607644C2 - Method of platinum group metals extracting from voloxidized snf acid dissolving product - Google Patents

Abstract

FIELD: fuel.

SUBSTANCE: invention relates to spent nuclear fuel (SNF) processing. Method of platinum group metals extracting from product of voloxidized SNF acid dissolving involves, that obtained after voloxidiation SNF is dissolved in nitric acid in the range of temperatures of 83–86 °C for 4–5 hours to obtain residual nitric acid content in product in range of 1.42–2.3 mol/l,uranium in range of 480–600 g/l, obtained product is thermostated in range of temperatures of 69–80°C for 2–48 hours, adding flocculant and dispersing reaction mixture, performing sediment accumulation in device bottom part by sedimentation deposition in range of temperatures of 35–57 °C for 6–24 hours. Product clarified part is separated by decantation, averaging sedimented deposit in remaining clarified solution amount in device.

EFFECT: invention enables to extract into deposit of more 78,7 % of platinoids and separate 98,1 % of suspensions formed during SNF dissolving.

15 cl, 2 ex

Description

The invention relates to the field of spent nuclear fuel reprocessing (SNF) and can be used in applied radiochemistry for the separation of platinum group metals from SNF, for the clarification of products of the acid dissolution of SNF, in hydrometallurgical technology for the isolation of platinoids with secondary precipitates from nitric acid solutions.

As a result of SNF reprocessing for acid dissolution operations, two technological streams containing platinoids are formed: insoluble precipitates after clarification of the initial product and clarified solution, which, after separation of uranium and plutonium, is transferred as a highly active raffinate at the stage of evaporation and vitrification. In the existing technology for radiochemical reprocessing of spent nuclear fuel, wastes containing platinoids are not processed and sent for disposal in forms, the subsequent extraction of valuable components from which is impossible.

In addition, the distribution of platinoids in the process chain creates a number of problems in the operations following dissolution. During the extraction process, platinoids, due to their ability to recover, form films of interfacial sediments and suspensions, thereby changing the distribution coefficients, the delamination rate, lead to the formation of microemulsions, which ultimately causes a violation of the operating mode of the extraction cascade. In the process of glass melting, platinoids form an independent phase with components of highly active waste (HLW), which reduces the quality of the matrix obtained for immobilizing HLW, unpredictably affects the process, changing the distribution of electric currents in the melting cooking zone.

Thus, the need to remove platinoids from SNF reprocessing products is caused both by technological expediency for the process of radiochemical reprocessing of SNF, and by obtaining valuable components in a concentrated and suitable form for further processing.

For the removal of platinoids from the spent nuclear fuel product, the most appropriate solution is the complete isolation of platinoids in the solid phase during dissolution-clarification operations. However, the use of specific selective precipitation methods for the isolation of platinoids from the product of the acid dissolution of SNF leads to a change in the conditions of the product itself and its unsuitability for extraction processing, which casts doubt on the feasibility of the proposed solution.

The use of quantitative isolation of platinoids by increasing their precipitation upon dissolution of non-oxidized SNF has not had a technological embodiment to date. Changes in the parameters of the dissolution process (temperature, solution acidity, dissolution time, etc.) and the subsequent formation of secondary precipitates does not lead to their stable extraction into the solid phase, and is also accompanied by an increase in the amount of plutonium removed from the precipitates.

An introduction to the technological chain of innovative redistribution - the operation of voloxidation significantly changes the composition of the formed precipitation and increases their amount. This is due to the formation of independent solid solutions of individual fragmentation elements in the main oxidized uranium-plutonium phase as a result of thermochemical treatment of spent nuclear fuel and the impossibility of their further dissolution in dilute solutions of nitric acid.

The prospect of the separation of metals of the platinum group (PGM) by the formation and separation of precipitation in the dissolution-clarification operations of the voloxidized SNF is a prerequisite for the development of this method.

The prior art method for producing concentrates of rhodium, palladium and ruthenium from nitric acid solutions [Patent RU 2239666, C22B 11/00, C22B 3/44, publ. 10.11.2004], which consists in the precipitation of platinoids from solution, including the introduction of nitric acid into the solution with a concentration of 2-3 mol / l in the temperature range 18-80 ° C of alkali metal thiocyanate in solid form with an excess of 1/3 of the amount of nitric acids. The disadvantages of the method are: the formation in solution of a sulfate ion incompatible with the subsequent extraction process as a result of oxidation of the thiocyanate ion; probability of coprecipitation of plutonium with sparingly soluble complex thiocyanate compounds of platinoids; the need for thermochemical decomposition of the resulting precipitation; the probability of radiocaesium inclusion in the composition of the formed complexes and, as a consequence, an increase in the gamma activity of precipitation.

A known method of extracting palladium from real highly active solutions [Radiochemistry, 2008, T. 50, No. 2, p. 151-153], which consists in the electrochemical deposition of palladium in the device of the original design using a short-circuited bimetallic pair, excluding the use of an external current source, at a nitric acid concentration of 0.5-3 mol / l and creating a potential from +0.4 to +0, 6 B. The disadvantages of the method are: the negative effect of elements of variable valency on the effectiveness of the reduction of palladium at the cathode from nitric acid solutions; a negative impact on the efficiency of recovery of palladium nitrous acid formed as a result of radiolysis and autocatalytic reactions; low cleaning factors associated with the difficulty of washing the precipitation.

The prior art method for the extraction of palladium from nitric acid solutions [Patent RU 2228380, C22B 011/00, C22B 003/44, publ. 10.12.2003], which consists in the fact that an oxidizing agent or a reducing agent is introduced into the initial nitric acid solution, and carbon monoxide or its compound with copper monochloride is used as a precipitant. The disadvantages of the method are: the inability to allocate palladium from the product of the acid dissolution of SNF due to the introduction of compounds that change the distribution coefficients during subsequent extraction processing; the introduction of chlorine compounds; the introduction of reagents that increase the amount of solid waste (as a result of the introduction of potassium permanganate), the absence of data confirming the selectivity of the process; lack of data on the effectiveness of testing the method on real products.

A known method for the isolation of platinoids with precipitates from the product of the dissolution of SNF [Radiochemistry, 2010, T. 52, No. 4, p. 342-345], which consists in studying the composition of the sediments in order to select the conditions for the dissolution of SNF and the conditions for the formation of secondary sediments, including dissolving non-oxidized fuel at a temperature of 80-95 ° C to obtain a nitric acid concentration of 1.7-3.5 mol / l in the product , uranium concentration of 250-350 g / l, the introduction of a flocculant, an exposure of 10-12 hours, the separation of primary sediments during filtration through a septum of 1-2.5 microns and 1-3.5 microns, the formation of secondary precipitates at a temperature of 20-80 ° C for 7-60 days, re-filtration. With primary precipitation, in this way, it is possible to extract Pd - 6.7-20.0%, Ru - 15.0-33.0%, Rh - 15.0-69.0%, the yield of platinoids in secondary precipitation is 1-2% from the initial content in SNF. The disadvantages of the method are: an insignificant amount of platinoids total separated; the lack of a stable result when withdrawing platinoids with primary sediments; the duration of the operation of the formation of secondary sediments, the inefficiency of separation of platinoids with secondary sediments, the coprecipitation of plutonium with secondary sediments.

The prior art method for the deposition of platinoids from aqueous solutions in the processing of spent nuclear fuel [Patent RU 2147619, C22B 011/00, C22B 003/44, C22B 007/00, publ. 04/20/2000], taken as a prototype, including introducing into the nitric acid solution with an acid concentration of 0.01-1.4 mol / L at a temperature of 70-80 ° C an excess of the sodium salt of formaldehyde sulfoxyl acid, carrying out the deposition and separation of the precipitate containing platinoids, providing separation of more than 99% of platinoids with a coefficient of purification from associated metals an average of 10 ³ . The disadvantages of the method are: the lack of testing data on real SNF processing products; the need to neutralize the product of the acid dissolution of SNF to obtain the claimed acidity range; an increase in the number of actinides removed with precipitates as a result of their coprecipitation with compounds of the claimed class; the need for oxidative stabilization of valence forms of actinides before extraction processing; the formation of compounds incompatible with the Purex process, the formation of a sulfate ion in the process of evaporation of the raffinate. In fact, the authors of the invention determined only a set of conditions under which platinoids can be isolated from solutions; the macro- and microcomposition of the actual spent fuel processing product are not taken into account. In addition, the corrosive effects of the introduced reagent and its degradation products on the structural materials of the equipment are unknown.

The objective of the present invention is to develop a method for the separation of platinoids from voloxidized spent nuclear fuel, which includes purification from fission products and concentration of valuable components, ensuring their isolation in a compact form, technologically suitable for implementation under conditions of radiochemical production, which meets the requirements of radiation safety and eliminates the formation of additional waste.

The problem is solved by increasing the completeness of the extraction of platinoids from the product of the acid dissolution of the voloxidized SNF due to the formation of secondary sediments directly on the primary and by simplifying the operation of separating the sediments.

The proposed method allows us to simultaneously solve the problem of clarification of the products of acid dissolution of SNF and obtaining conditional in terms of solid phase content of products of acid dissolution of SNF sent to further extraction processing.

The technical result of the invention is the precipitation of more than 78.7% of platinoids and the separation of 98.1% of the suspension formed during the dissolution of SNF. Achieving the technical result ensures a decrease in the number of interfacial suspensions and stabilized emulsions during extraction processing, a decrease in the amount of precipitation during evaporation of the raffinates, and the formation of an independent metal phase during vitrification of HLW.

To achieve a technical result in a method for the extraction of platinum group metals from an acid dissolution product of a voloxidized spent nuclear fuel, the SNF obtained after the oxidation of SNF is dissolved in nitric acid in the temperature range 83-86 ° C for 4-5 hours to obtain a residual nitric acid content in the product in the range 1.42-2.3 mol / l, uranium in the range of 480-600 g / l, the product is thermostated in the temperature range of 69-80 ° C for 2-48 hours, a flocculant is introduced and the reaction mixture is dispersed, wire t accumulation of sludge in the bottom of the apparatus due to sedimentary deposition 35-57 ° C temperature range for 6-24 hours, the product was separated by decanting the clarified portion averaged sedimented precipitate remaining in the clarified solution volume in the vehicle, the precipitate is separated from the product.

An increase in the reactivity of powdered voloxidized SNF is achieved by pretreatment with vigorous stirring with a solution of hydrogen peroxide with a concentration of 20-90 mg / l at a mass ratio of T: W = 1: 4-1: 20 in the temperature range 25-40 ° C, decantation 95% solution, washing and decanting a 90% washing solution, subsequent processing with vigorous stirring with a weak alkali solution with a concentration of 4.2-10.3 mmol / l at a mass ratio of T: W = 1: 3-1: 17 in the temperature range 25-95 ° C, decantation of 90% solution and prom Coy distillate decanting 90% of the wash solution.

An increase in the amount of precipitating elements precipitated during the dissolution of voloxidized SNF is achieved by using nitric acid with a concentration of 12 mol / L at a temperature of 83-86 ° C, reducing the amount of (calculated) nitric acid taken for dissolution, with obtaining residual acidity in the product 1.42-2.3 mol / l, increasing the concentration of uranium in the product of the acid dissolution of SNF to 480-600 g / l.

A decrease in the amount of uranium that is precipitated to a level of less than 0.13% and plutonium to a level of less than 0.6% is achieved by increasing the dissolution time to 4-5 hours and the temperature regime (temperature range 69-80 ° C) of the subsequent thermostating of the spent fuel acid dissolution product .

An increase in the content of hydrated forms of platinoids to the level necessary for the formation of secondary precipitation is provided by a decrease in the residual content of nitric acid in the product (to 1.42-2.3 mol / L or less), by thermostating in the closed system of the obtained product at a temperature of 69-80 ° C within 2-48 hours and is due to the ability of platinoids to form the most stable forms of mixed polynuclear nitrite aquacomplexes in the presence of a radiolitically generated nitrous acid in the solution.

The separation of PGM compounds into secondary precipitates is based on a decrease in the solubility of their polynuclear aquacomplexes modified with hydroxyalkylated amines and further precipitation of the latter on the surface of the solid phase of primary precipitates due to an increase in the polymerization degree of platinum nitrite aquacomplexes by coordinating the inclusion of an oxyalkylated amine-containing polymer in the complex structure. The precipitation of platinoids in the precipitate is achieved by introducing into the product an acid dissolution of SNF flocculant, which includes one or more amino groups, and its subsequent dispersion, thermostating of the product in the temperature range of 35-57 ° C for 6-24 hours. In a particular case, diproxamine-157 with a concentration of 100-250 mg / l is used as a flocculant.

Clarification of the product of the acid dissolution of SNF is due to enlargement of solid particles to a size of 3-20 μm under the influence of a flocculant and is achieved by quantitative sedimentation of the formed precipitate with a temperature of 22-24 hours. The solids content in the clarified product does not exceed 1.6-1.9% of the initial content and averages 220-390 mg / l. Obtaining dense easily sedimented particles of the solid phase is achieved by using primary precipitation as a collector for the formation of secondary precipitation. The formation of secondary precipitation is carried out without separation of primary precipitation. An indirect indicator of the completeness of the removal of platinoids in the precipitate is an increase in the mass of the solid phase due to the formation of secondary precipitates, comprising 33.1-50.2% of the mass of primary precipitation and 0.76-1.5% of the mass of fuel. An indirect indicator of the compactness of sedimented sediment is its volume fraction, which is no more than 1-5% of the product volume.

The total amount of separated sediment is 2.3-3.1% of the mass of spent nuclear fuel.

The compactness and density of the resulting precipitation provides the possibility of decantation of at least 87% vol. clarified solution. The content in the decantate of sediment-forming components and a solid phase corresponds to the required conditions for a stable feed stream of the extraction redistribution and does not require additional clarification operations. The precipitate is washed out in the remaining volume of the clarified solution using known methods (pulsating water flow or a stream of bubbling air, rotational mixing, ultrasonic dispersion), the resulting solid phase concentrate is transferred to clarification. The completeness of clarification is ensured by the enlargement of particles of the solid phase as a result of the coagulation effect on the product of the acid dissolution of SNF during the formation of secondary sediments.

The coefficients of purification of precipitated platinum group metals from gamma-active fission products are 2.3 × 10 ³ -1.3 × 10 ⁴ .

Example 1

The bulk material obtained after the oxidation wave (VVER-1000 SNF with a burn-up depth of 51.86 GW · day / tU after 10 years of exposure) in the amount of 34.1 g was subjected to activation: 70 ml of a solution of hydrogen peroxide with a concentration of 20 mg / l were treated at a temperature 30 ° C and stirring at a stirring device rotation speed of 800 rpm for 15 minutes, the solid phase was defended for 2 hours, 63 ml of the solution was decanted, 200 ml of distillate was washed for 15 minutes with vigorous stirring, and the solid phase was defended for 2 hours, decanted 195 l of the washing solution, the solid phase was treated with 55 ml of a solution of sodium hydroxide with a concentration of 4.4 mmol / l at 35 ° C with vigorous stirring, the solid phase was lagged for 2 hours, 53 ml of the solution was decanted, three times washing with 100 ml of distillate was carried out with settling for 2 hours and decantation of the wash solution. Wash and working solutions were filtered on a microfiltration partition MFPK-1G with a pore size of 150 nm. The amount of solid phase removed with decantates was 0.03% (by dry residue) of the mass of spent nuclear fuel.

A calculated amount of nitric acid was added to a device with a portion of prepared activated voloxidized SNF to a concentration of 12 mol / L. Dissolution was carried out for 4 hours at a temperature of 85 ± 0.5 ° C with vapor condensation in a reflux condenser at a temperature of 1-2 ° C.

The concentration of uranium in the product of the acid dissolution of SNF was 509.5 ± 4.2 g / l, the concentration of nitric acid was 95.9 g / l. The volume of the product was 58.6 ± 0.3 ml. The concentration of the solid phase in the product was 9.9 ± 0.4 g / l (taking into account only primary precipitation), the amount of primary precipitation was 1.55% of the mass of spent nuclear fuel.

The obtained product of the acid dissolution of SNF was kept at a temperature of 80 ± 0.5 ° C for 42 hours, after which a solution of diproxamine-157 was added to the product with vigorous stirring (with a rotary stirrer at a speed of 800 rpm) to obtain a flocculant concentration in the product of 250 mg / l The resulting solution was thermostated at a temperature of 48 ± 0.5 ° C for 22 hours.

The concentration of the solid phase (after the formation of secondary sediments) in the product before separation of the precipitate was 19.7 ± 0.4 g / l.

47.6 ± 0.3 ml of clarified product was separated by decantation. The precipitate was averaged in the remaining volume of the product and separated on a microfiltration partition of MFK-OG with a pore size of 50 nm, washed three times with 50 ml of a solution of nitric acid with a concentration of 0.5 mol / L and 50 ml of distillate. The washed precipitate was dried at a temperature of 90 ° C for 30 hours, the mass was determined, and it was completely opened in a rigid oxidizing system. The solution after opening the precipitate was analyzed for the platinum content by an inductively coupled plasma spectrometric method (ICP spectrometer), the uranium content was determined spectrophotometrically, the plutonium content was determined by the alpha radiometric method, the specific gamma activity of the precipitate was determined in the solution from the opening of the precipitate by the gamma spectrometric method. The content of valuable components in the sediment was: U - 0.13%, Pu - 0.6%, Ru - 94.0%, Rh - 97.3%, Pd - 97.4%, Ag - 14.9%. The coefficient of purification of PGMs precipitated from gamma-active fission products was 2.3 × 10 ³ . The amount of separated sediment was 3.04% (per dry weight) of the mass taken to dissolve SNF, the amount of secondary sediment was 49.1% of the total mass of precipitation. The proportion of the introduced flocculant in the sediment did not exceed 1.3%.

The amount of solid phase in the clarified product was 1.61% of the initial amount. The specified method of clarification of the product of the acid dissolution of SNF was used to organize a steady supply flow of extraction redistribution and excluded the formation of interfacial suspensions, ensuring stable operation of the equipment.

Example 2

The bulk material obtained after voloxidation (SNF VVER-1000 with a burnup depth of 51.86 GW · day / tU after 10 years of exposure) in the amount of 34.8 g was not subjected to activation, was dissolved analogously to example 1.

The concentration of uranium in the product of the acid dissolution of SNF was 504.5 ± 4.0 g / l, the concentration of nitric acid was 125.1 g / l. The volume of the product was 59.4 ± 0.3 ml. The concentration of the solid phase in the product was 9.0 ± 0.4 g / l (taking into account only primary precipitation), the amount of primary precipitation was 1.53% of the mass of spent nuclear fuel.

The obtained product of the acid dissolution of SNF was kept at a temperature of 77 ± 0.5 ° C for 2 hours, after which a solution of diproxamine-157 was added to the product with vigorous stirring (with a rotary stirrer at a speed of 300 rpm) to obtain a flocculant concentration in the product of 150 mg / l The resulting solution was thermostated at a temperature of 53 ± 0.5 ° C for 18 hours.

The concentration of the solid phase (after the formation of secondary sediments) in the product before separation of the precipitate was 14.5 ± 0.4 g / l. The separation of the precipitate and analysis of the solid phase was carried out analogously to example 1.

The content of valuable components in the sediment was: U - 0.09%, Pu - 0.18%, Ru - 91.3%), Rh - 77.8%, Pd - 97.5%, Ag - 27.7%. The coefficient of purification of precipitated platinoids from gamma-active fission products was 3.9 × 10 ³ . The amount of separated sludge was 2.31% (by weight of dry residue) of the mass taken to dissolve SNF, the amount of secondary sludge was 33.8% of the total mass of precipitation. The proportion of introduced flocculant in the sediment did not exceed 1.7%.

The amount of solid phase in the clarified product was 1.85% of the initial amount. The specified method of clarification of the product of the acid dissolution of SNF was used to organize a steady supply flow of extraction redistribution and excluded the formation of interfacial suspensions, ensuring stable operation of the equipment.

The essence of the invention is to create conditions for the isolation of platinoids from the products of acid dissolution of SNF before extraction processing by increasing the content of insoluble forms of platinoids in primary sediments, increasing the overall completeness of the isolation of platinoids by separating their soluble forms with secondary sediments, enlarging particles of the formed solid phase, simplifying it subsequent separation.

When developing a method for separating platinoids from SNF, the main goal is not to complete the selection of the target component, but to ensure nuclear and radiation safety requirements, to eliminate the formation of additional waste, to minimize the impact of the proposed process on subsequent operations of processing a uranium-containing product as part of the general concept of the technological process. Therefore, the proposed method has the following advantages over the prototype: the minimum number of changes to the process at the operations of dissolution-clarification; exclusion of restorative effects on the product; the exclusion of the introduction of salt-forming agents and corrosive additives into the product; organization of the process of extraction of platinoids on the real product of the acid dissolution of SNF; the suitability of the clarified product for further radiochemical processing.

Claims (15)

1. The method of extraction of platinum group metals from the product of the acid dissolution of voloxidized SNF, characterized in that the obtained after voloxidation of SNF is dissolved in nitric acid in the temperature range 83-86 ° C for 4-5 hours to obtain a residual content of nitric acid in the product in the range 1.42-2.3 mol / l, uranium in the range of 480-600 g / l, the product obtained is thermostated in the temperature range of 69-80 ° C for 2-48 hours, a flocculant is introduced and the reaction mixture is dispersed, and sediment is accumulated in the bottom of the apparatus for Thu sediment deposition in the temperature range of 35-57 ° C for 6-24 hours, the product was separated by decanting the clarified portion averaged sedimented precipitate remaining in the clarified solution volume in the vehicle, the precipitate is separated from the product.  2. The method according to p. 1, characterized in that before dissolving the voloxidized SNF is subjected to separate treatment with a solution of hydrogen peroxide with a concentration of 20-90 mg / l in the temperature range of 25-40 ° C and an alkali solution with a concentration of 4.2-10.3 mmol / l in the temperature range 25-95 ° C. 3. The method according to p. 2, characterized in that the mass ratio of voloxidized spent fuel to the amount of hydrogen peroxide solution is defined as T: W = 1: 4-1: 20. 4. The method according to p. 2, characterized in that the mass ratio of voloxidized spent fuel to the amount of alkali solution is determined as T: W = 1: 3-1: 17. 5. The method according to p. 1, characterized in that when dissolving voloxidized spent nuclear fuel use nitric acid with a concentration of 12 mol / L. 6. The method according to p. 1, characterized in that the flocculant used contains in its composition one or more amino groups. 7. The method according to p. 1, characterized in that the flocculant used contains one or more oxyalkylated amino groups. 8. The method according to p. 1, characterized in that as a flocculant use diproxamine-157 at a concentration of 100-250 mg / L. 9. The method according to p. 1, characterized in that when decanting the product, at least 87% vol. clarified solution. 10. The method according to p. 1, characterized in that the formation of secondary precipitation is carried out without separation of the primary precipitation. 11. The method according to p. 1, characterized in that the total amount of separated sediment is 2.3-3.1% by weight of spent fuel. 12. The method according to p. 1, characterized in that the amount of secondary precipitation is 33.1-50.2% of the total mass of precipitation. 13. The method according to p. 1, characterized in that the amount of precipitated uranium does not exceed 0.13%, the amount of plutonium precipitated does not exceed 0.6%. 14. The method according to p. 1, characterized in that in the clarified product of the acid dissolution of SNF, the solids content does not exceed 1.6-1.9% of the initial content in the product. 15. The method according to p. 1, characterized in that the coefficients of purification of precipitated platinum group metals from gamma-active fission products are 2.3 × 10 ³ -1.3 × 10 ⁴ .