Two isomeric monoamides, dioctyl butyramide (DOBA) and dioctyl isobutyramide (DOIBA) were synthesized for extracting uranium(VI) and plutonium(IV) from aqueous nitric acid medium into various diluents such asn-dodecane, tertiary butyl benzene and xylene. DOBA extracted uranium(VI) and plutonium(IV) efficiently whereas DOIBA extracted uranium(VI) with negligible extraction for plutonium(IV). Both these cations were extracted as their disolvates. The thermodynamic parameters involved in the extraction determined by the temperature variation method indicated the reactions in all cases to be enthalpy favoured and entropy disfavoured. Possibility of separating micrograms of plutonium(IV) from macroquantities of uranium(VI) using the mixture of these amides was explored.
The unsymmetrical diamide methylbuthylmalonamide has been synthesized and used in the extraction of U(VI), Pu(IV) and Am(III) in benzene medium. The distribution ratio for the three cations was found to increase with increasing aqueous nitric acid concentration. U(VI) and Pu(IV) were found to be extracted as disolvates while Am(III) as a trisolvate. The thermodynamic parameters determined by the temperature variation method showed the extraction reactions to be mainly enthalpy-controlled. Am(III) was found to be back-extracted with dilute nitric acid, while Pu(IV) by dilute nitric acid-hydrofluoric acid mixture and U(VI) by dilute sodium carbonate solution.
The extraction of uranium(VI) and plutonium(IV) was carried out with two isomeric monoamides, dihexylbutyramide (DHBA) and dihexylisobutyramide (DHIBA) from nitric acid medium, usingn-dodecane as diluent. The possibility of separation of the two metal ions from each other without valency adjustment was attempted. U(VI) was extracted as its disolvate, while Pu(IV) was extracted as its trisolvate. From the variation of distribution ratio with temperature, it was shown that the extraction reaction was enthalpy controlled in all the cases.
The extraction behavior of U(VI) and Pu(IV) with dioctyloctanamide (DOOA), dioctylethylhexanamide (DOEHA) and diisobutylethylhexanamide (DIBEHA) was investigated from nitric acid medium. With DOOA, U(VI) extraction is higher than that for Pu(IV) upto 5M HNO3 and the trend is reversed at higher acid concentrations. Extraction yield of U(VI) is higher than that for Pu(IV) in the case of DOEHA and DIBEHA. DIBEHA extraction of Pu(IV) is found to be very small. The lower value of the distribution ratio for Pu(IV) with branched amides was attributed to steric reasons. The possibility of using these amides for separation of U(VI) and Pu(IV) without valency adjustment was explored. Both U(VI) and Pu(IV) are extracted as their disolvates by DOOA and DOEHA.
The extraction of uranium(VI) and plutonium(IV) from nitric acid into n-dodecane was studied using two isomeric branched alkyl amides, di(2-ethyl hexyl) butyramide (DEHBA) and di(2-ethyl hexyl) isobutyramide (DEHIBA). The extraction ratios of Pu(IV) at relatively high acidities were higher than the corresponding values for U(VI) in the case of DEHBA. However, with DEHIBA the values for Pu(IV) were negligibly small. Pu(IV) was found to be extracted as trisolvate by DEHBA and as disolvate by DEHIBA. U(VI) was extracted by both the amides. From the study of the extraction reactions at different temperatures, it was shown that all the reactions in the present investigation were enthalpy favoured and entropy disfavoured. Separation of Pu(IV) from bulk of U(VI) was feasible. However, the purity of the separated plutonium was not satisfactory in batch extraction studies.
The stability constants of the complex of Pu/III/ with a humic acid at pH 2.9 and 5.0 were measured using solvent-extraction technique. The organic extractant was dinonyl naphthalene sulphonic acid in sodium form /NaD/ in benzene while the humate aqueous phase had a constant ionic strength of 0.5M /NaClO4/. The total carboxylate capacity of the humic acid was determined by direct potentiometric titration to be 6.201 Meq g–1. The apparent pKa increased as the degree of ionization // increased. The 1g
1 values of the complex of Pu/III/ with humic acid have been determined to be 2.8 and 3.11 at pH 2.9 and 5.0, respectively.
An improved method for the determination of plutonium in an aliquot using cerium/IV/ as an oxidant is reported. Plutonium is oxidized quantitatively to plutonium/VI/ in nitric acid medium by cerium/IV/, the excess of which is chemically destroyed in a single step by hydrochloric acid. Plutonium/VI/ is then reduced to plutonium/IV/ with a known amount of Fe/II/, the excess of which is back titrated potentiometrically with standard dichromate. Results of analysis of 3–5 mg amounts of plutonium in aliquots containing standard plutonium nitrate solution are reliable within 0.2%. Effect of the presence of some relevant foreign ions has been studied. The application of the method for the analysis of mixtures containing various amounts of uranium and plutonium has been examined.
The liquid-liquid extraction behavior of plutonium(IV) from aqueous nitric acid media into n-dodecane by di(2-ethylhexyl)sulfoxide (DEHSO) was investigated over a wide range of conditions. Optimum-parameters such as the aqueous phase acidity, reagent and metal concentrations, etc., were established for efficient extraction-separation of tracer as well as macro levels of plutonium. It was found that the extraction increased with increasing nitric acid concentration up to 6M HNO3 and then decreased. Extraction also increased with increasing extractant concentration. After loading of the organic phase with 2 to 50 mg/ml of U(VI), extractability of Pu(IV) became considerably lower. Recovery of Pu(IV) from the organic phase was accomplished using dilute uranium(IV) nitrate as the strippant.
A quartz pyrohydrolysis apparatus designed for the safe handling of pyrophoric carbide nuclear fuel samples while separating C1 and F trace impurities for quality control analysis is described. It has several advantages over the commonly used pyrohydrolysis apparatus. Performance of the apparatus during routine analysis of a large number of nuclear fuel samples has been found to be quite satisfactory.