[en] To increase capacity of extractant for solvates of Th and Pu nitric acid salts it is suggested to use additional methyl-derivatives of paraffin hydrocarbons C₁₀-C₁₆ in extractant, containing TBP and n-paraffin hydrocarbons. Ratio of components: TBP-30-50%; methyl-derivatives of paraffin hydrocarbons-20-65%; n-paraffin hydrocarbons - the rest. Extractant capacity with respect to Th(4) can be increased ∼2,4 times, and with respect to Pu(4)-1.4 times

[en] A study was made on the dependence of the rate of Pu⁶⁺ macroconcentration reduction to Pu⁴⁺ (2-40 g/l) when bubbling mixture of gaseous nitrogen oxides (oxide and dioxide) on nitric acid concentration (3-8 mol/l, initial Pu⁶⁺ content (20-95 %) total plutonium concentration in processable solution temperature (50-80 deg C), the rate of gaseous nitrogen oxides and composition of the given gaseous mixture, as well as the presence of uranium (6) and stable isotopes of some fission products in processable solution. Empirical kinetic equation for the rate of studied process enabling to calculate the apparent rate constants of gas-chemical reduction of Pu⁶⁺ to Pu⁴⁺ using experimental data is suggested. Applicability of derived equation for decribing the studied process under steady conditions during regeneration of NPP spent fuel is shown

[en] Method for decreasing plutonium losses during Pu(6) to Pu(4) reduction by nitrogen oxides in nitric acid solutions at temperatures above 60 deg C is suggested. Nitric acid solution (4-5 mol/l), saturated preliminary by nitrogen oxides at 22-25 deg C is used as source of nitrogen oxides at that. This results to sufficient decrease of plutonium losses with aerosol and 5-30 time decrease of the flow-rate of gaseous nitrogen oxides. The method can be used for reprocessing of plutonium-containing materials and for reprocessing of NPP spent fuel in particular

[en] Short note

[en] While acknowledging the bold and persistent efforts of U.S. and Russian specialists to develop the concept of pyrochemical reprocessing of spent nuclear fuel from fast reactors on remote-controlled equipment for removal of actinides from the fission products one should recognize that the tasks of reprocessing such fuel can be handled only by using water-extraction technology, especially since the known Purex process continues to be improved to the point that a single-cycle scheme may be developed. This article presents results of pilot studies conducted in hot cells using multistage extractors in continuous counterflow operation; data on various extractor types used in reprocessing spent mixed oxide nuclear fuel; advantages and disadvantages of centrifugal and pulsed column extractor; comparison of column-type and centrifugal extractors; and extraction process