The equilibrium constant for the first hydrolysis reaction of tetravalent plutonium is surrounded by uncertainty. A new method
illustrates criteria by which the reliabilities of the numerical estimates can be judged. The new formulas are simple, the
method is easy to apply, and the results are easy to compare.
The numerical value of the first hydrolysis constant of tetravalent plutonium is uncertain by a factor of about ten. This
article illustrates the estimation of that constant by a least squares method applied to simultaneous equations involving
all of the Pu oxidation states.
Boundary equations for a diagram of the ambiguous, forbidden, and unique combinations of hexavalent plutonium versus the Pu
oxidation number are presented. The equations and the diagram apply after disproportionation reactions have reached equilibrium.
A curve illustrates the equilibrium fraction of Pu(V) versus the Pu oxidation number in a solution at pH 2.
Authors:P. Govindan, S. Sukumar, K. Vijayan, G. Santhosh Kumar, S. Ganesh, Pradeep Sharma, K. Dhamodharan, R. Subba Rao, M. Venkataraman, and R. Natarajan
A novel method has been developed for recovery of plutonium and uranium from carbonate wash solutions generated during solvent
wash process involved in the reprocessing of high burn up FBTR fuel. The proposed method involves a selective coprecipitation
of Pu and U by adding ammonium hydroxide to the pre acidified carbonate wash solution. Substantial removal of DBP by successive
steps of coprecipitation, completely eliminates the possibility of undesired solid formation which is mainly due to the presence
of high content of DBP. By adopting this method, an excellent decontamination factor for DBP has been achieved without any
crud/solid formation. Phosphate content in the final oxide product meets the product specifications. Flowsheet condition necessary
for the recovery process for plutonium from the aqueous carbonate solution is formulated and adopted in the CORAL facility.
Authors:V. Rao, I. Pius, M. Subbarao, A. Chinnusamy, and P. Natarajan
A method for the precipitation of plutonium(IV) oxalate from homogeneous solutions using diethyl oxalate is reported. The precipitate obtained is crystalline and easily filterable with yields in the range of 92–98% for precipitations involving a few mg to g quantities of plutonium. Decontamination factors for common impurities such as U(VI), Am(III) and Fe(III) were determined. TGA and chemical analysis of the compound indicate its composition as Pu(C2O4)2·6H2O. Data are obtained on the solubility of the oxalate in nitric acid and in mixtures of nitric acid and oxalic acid of varying concentrations. Green PuO2 obtained by calcination of the oxalate has specifications within the recommended values for trace foreign substances such as chlorine, fluorine, carbon and nitrogen.
Determination of actinides in the environmental and bioassay samples is important in view of the following factors: increasing energy production by nuclear reactors; environmental contamination due to fallout from nuclear weapons testing and burn up of nuclear-powered satellites; the growing emphasis on the desirability of a cleaner environment; and public concern over the potential hazards associated with nuclear reactors. Among the various actinides, plutonium is one of the most important due to the large amounts produced in the nuclear fuel cycle. Further, the extremely low levels of plutonium in the different biological and environmental samples demand the development of precise, accurate, and sensitive methods to arrive at meaningful conclusions from the results obtained in various studies. In addition to various other techniques available, alpha spectrometry is commonly used.
A stable green solution of tricarbonatocobaltate(III) has been prepared and used for the redox titrimetric determination of plutonium in HNO3 medium. Quantitative oxidation could be achieved and excess oxidant could be destroyed by NaNO2. Pu(VI) was deter-ined by adding known excess of Fe(II) and carrying out potentiometric titration. The precision at the level of 0.5–5.0 mg was 2% RSD.
The work of disproportionation of tetravalent plutonium can be estimated in two ways. In the first way, the Pu4+cation generates three other oxidation states and PuOH3+. In the second way, the sum of Pu4+and PuOH3+generates three other oxidation states and fraction of PuOH3+changes. The methods yield different estimates of the work of Pu4+disproportionation.
Authors:Silvia Dulanská, Boris Remenec, L’ubomír Mátel, and Erik Durkot
A rapid separation method has been developed which allows measurement of plutonium, americium and strontium isotopes in the
radioactive sludge from Nuclear power plant A1 Jaslovske Bohunice (NPP A1) with high chemical recoveries and effective removal
of matrix interferences. This method uses different commercial products stacked AnaLig® Pu02, AnaLig® Sr01 and TRU® Resin cartridges from IBC Advanced Technologies and Eichrom Technologies. The method allows the rapid separation of plutonium,
strontium and americium using a single multi-stage column in the vacuum box (cartridge technology) with rapid flow rates to
minimize sample separation time. The 239,240Pu, 238Pu and 241Am were determined by alpha spectroscopy, 90Sr was counted on TRICARB 2900 TR by Cerenkov counting of its progeny 90Y.
Authors:P. Pathak, D. Prabhu, M. Bindu, S. Tripathi, and V. Manchanda
The primary purpose of this study was to understand the alpha radiolytic degradation behavior of N,N-dihexyl octanamide (DHOA) vis a vis tributyl phosphate (TBP) solutions in n-dodecane under plutonium loading conditions. These studies were carried out as a function of dose on different Pu loaded
samples (containing 0.002-10 g/L Pu) from 4 M HNO3 medium. These Pu loaded solutions were evaluated for stripping behavior by contacting with 0.5 M NH2OH at 0.5 M HNO3 solutions. Organic phase analysis was carried out by gas chromatography (GC) and by visible spectrophotometry. These studies
clearly indicated that Pu stripping becomes difficult with increased dose in the case of TBP system. On the other hand, no
such problem was observed in DHOA system during stripping of plutonium, thereby indicating that DHOA is a promising candidate
for the reprocessing of high burn up Pu rich spent fuels.