T. I. Trofimov

Recovery of U and Pu from simulated spent nuclear fuel by adducts of organic reagents with HNO3 followed by their separation from fission products by countercurrent chromatography

Abstract The primary patterns of isolation of actinides from their dioxides with supercritical carbon dioxide containing adducts of tri-n-butyl phosphate and other organic reagents with nitric acid has been investigated. The effective extraction of actinides from their dioxides was achieved using these adducts under common (ambient) conditions without supercritical carbon dioxide, which makes the performance of the process significantly easier. Extraction of uranium and the primary fission products from a simulated spent nuclear fuel has been studied. It was shown that at almost quantitative extraction of U (>99%), most 95Zr and 95Nb (up to 90%) as well as 131I, partially 103Ru (∼30%), 99Tc (20%) and lanthanides (up to 10%) also passed into the adduct phase. In this case complete separation of U from 137Cs, 85Sr, 140Ba as well as from the bulk of 140La, 141–144Ce and 147Nd was observed. After dissolution of the adduct containing U and partially fission products in liquid carbon dioxide the extract was transported into a separation column of a planetary centrifuge, and separation of U from remaining fission products was performed by countercurrent chromatography in the dynamic mode.

New approaches to reprocessing of oxide nuclear fuel

Dissolution of UO2, U3O8, and solid solutions of actinides in UO2, including those containing Cs, Sr, and Tc, in weakly acidic (pH 0.9–1.4) aqueous solutions of Fe(III) nitrate or chloride was studied. Complete dissolution of the oxides is attained at a molar ratio of Fe(III) nitrate or chloride to uranium of 1.6 or 2.0, respectively. In the process, actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). At 60°C, actinide oxides dissolve in these media faster than at room temperature. In the solutions obtained, U(VI) and Pu(III) are stable both at room temperature and at elevated temperatures (60°C), and also at high U concentrations (up to 300 mg ml−1) typical of process solutions (6–8 M HNO3, ∼60–80°C). After the oxide fuel dissolution, U and Pu are recovered from the solution by peroxide precipitation. In so doing, the content of Fe, Tc, Cs, and Sr in the precipitate does not exceed ∼0.05 wt %. From the solution after the U and Pu separation, the fission lanthanides, Tc, Cs, and Sr can be recovered by precipitation of Fe hydroxides in the presence of ferrocyanide ions and can be immobilized in appropriate matrices suitable for long-term and environmentally safe storage.

Dissolution of Actinide Oxides in Supercritical Fluid Carbon Dioxide, Containing Various Organic Ligands

The dissolution of actinide oxides in supercritical fluid carbon dioxide containing a complex of tri-n-buty 1 phosphate with nitric acid was investigated. It was shown for the first time that milligram amounts of uranium dioxide can be quantitatively dissolved in supercritical carbon dioxide containing this reagent and efficiently separated from plutonium, neptunium, and thorium on its supercritical fluid extraction from a mixture of their oxides. The quantitative dissolution of milligram amounts of uranium trioxide in supercritical carbon dioxide containing thenoyltrifluoroacetone and tri-n-butyl phosphate was first performed using ultrasonication.

Use of microwave radiation for preparing uranium oxides from its compounds

The formation of uranium oxides by thermal decomposition of uranyl diaquadihydroxylaminate monohydrate, ammonium diuranate, ammonium tricarbonatouranylate, and uranium peroxide under the action of microwave (MW) radiation was studied. Uranium dioxide is formed by decomposition of these compounds in a reducing atmosphere at the MW radiation power of 600 W and treatment time of 5–10 min. In air, under the same conditions, U3O8 is formed. Under the action of MW radiation, substandard ceramic pellets of UO2 fuel can be readily converted in air to powdered U3O8. The use of MW radiation for thermal decomposition of uranium compounds allows the power and time consumption to be considerably reduced relative to the process with electrical resistance furnaces. A quick method for gravimetric testing of the composition of uranium oxides (UO2 or U3O8) using MW radiation was suggested.

Preparation of Np, Pu, and U dioxides in nitric acid solutions in the presence of hydrazine hydrate

Heating of nitric acid solutions of Np and Pu (∼90°C) in the presence of hydrazine hydrate (HH) leads to the formation of their hydrated dioxides in solution, transforming into crystalline dioxides at 300°C. Thermolysis of a mixed solution of U, Np, and Pu nitrates under the same conditions initially yields hydrated (U,Np,Pu)O2·nH2O, which on heating in air to ∼300°C transforms into a crystalline solid solution of (U,Np,Pu)O2. This method for stabilization of U dioxide in the presence of Pu in an oxidizing atmosphere can be used for preparing (U,Pu)O2 solid solutions of variable composition. This procedure shows doubtless prospects as a simple, efficient, and relatively low-temperature method for the production of MOX fuel for fast reactors.

Behavior of fission products in the course of dissolution of simulated spent nuclear fuel in iron nitrate solutions and of recovery of uranium and plutonium from the resulting solutions

Experiments aimed to examine the spent nuclear fuel dissolution in iron(III) nitrate solutions and to elucidate the behavior of fission products in the process were performed with simulated fuel corresponding to spent nuclear fuel of a WWER-1000 reactor. In Fe(III) nitrate solutions, U is quantitatively transferred from the fuel together with Cs, Sr, Ba, Y, La, and Ce, whereas Mo, Tc, and Ru remain in the insoluble precipitate and do not pass into the solution, and Nd, Zr, and Pd pass into the solution to approximately 50%. The recovery of U or jointly U + Pu from the solution after the dissolution of oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.

Recovery of rare earth elements, uranium, and thorium from monazite concentrate by supercritical fluid extraction

Quantitative recovery of rare earth elements (REEs), Th, and U by supercritical fluid extraction (SCFE) with carbon dioxide containing adducts of TBP and HDEHP with HNO3 directly from monazite concentrate (MC) powder is impossible and requires the conversion of the constituent elements into more soluble compounds. Microwave (MW) radiation can be efficiently used for MC pretreatment by sintering with Na2CO3 in the presence of coal. The resulting product consists of two phases. One of them contains REEs (∼50%) recoverable by supercritical carbon dioxide (SC-CO2) containing adducts of TBP or HDEHP with HNO3. The second phase is a solid solution of CeO2 with Th and U oxides and remaining amount of REEs. It is resistant to SCFE. Conditions were determined for quantitative dissolution of this phase in a mixture of 4 M HCl with 0.05 M HF. The use of HDEHP under the SCFE conditions allows quantitative recovery of Th and U from the hydrochloric acid solution. In the process, REEs remain in the aqueous phase and are thus separated from Th and U. A possible flowsheet was suggested for the recovery REEs from MC using SCFE with their simultaneous separation from Th and U.

Preparation of powdered uranium oxides by microwave heating of substandard ceramic pellets of oxide nuclear fuel

Microwave (MW) heating of substandard ceramic UO2 pellets in air allows their rapid conversion into powdered U3O8, from which UO2 can be obtained again in a reducing atmosphere. Comparative analysis of the physicochemical and technological properties of the U3O8 and UO2 powders obtained under the action of MW radiation with the industrial (standard) powders demonstrated their suitability for fabricating fuel pellets. The power consumption for MW heating appears to be lower by an order of magnitude than the power consumption for performing similar operations with electric resistance furnaces.

Preparation of uranium oxides in nitric acid solutions by the reaction of uranyl nitrate with hydrazine hydrate

UO2·nH2O formed by thermal denitration of uranyl nitrate in solutions under the action of hydrazine hydrate can be converted in air to UO3 at 440°C and to U3O8 at 570–800°C, and also to UO2 in an inert or reducing atmosphere at 280–800°C. After the precipitation of hydrated uranium dioxide, evaporation of the mother liquor at 90°C in an air stream allows not only evaporation of water, but also complete breakdown and removal of hydrazine hydrate and NH4NO3. The use of microwave radiation considerably reduces the time required for complete thermal denitration of uranyl nitrate in aqueous solution to uranium dioxide, compared to common convective heating.

Factors governing the efficiency of dissolution of UO2 ceramic pellets in aqueous solutions of iron nitrate

Dissolution of ceramic UO2 in aqueous Fe(NO3)3 solutions at different temperatures under the conditions of limited contact with air and in the autoclave mode was studied. In the course of UO2 dissolution at 60–90°C, the U/Fe molar ratio appears to be ∼1, whereas at room temperature (25°C) this value is ∼0.5. By varying the acidity of Fe nitrate solutions at these temperatures, it is possible to increase the U/Fe molar ratio to ∼4 and to obtain uranyl nitrate solutions with simultaneous removal of Fe from the solution in the form of a precipitate of the basic salt, or to perform quantitative dissolution of UO2 under the conditions excluding the formation of such precipitate. In the course of dissolution of ceramic UO2 in Fe(NO3)3 solutions, the appearance or absence of Fe(II) ions, the formation or absence of the precipitate of the Fe basic salt, and variation of solution pH are interrelated and are determined by the process temperature.