Efficacy of 3,4,3-LI(1,2-HOPO) for decorporation of Pu, Am and U from rats injected intramuscularly with high-fired particles of MOX

This study aimed to assess the efficacy of 3,4,3-LI(1,2-HOPO) for reducing uranium, plutonium and americium in rats after intramuscular injection of (U-Pu)O2 particles (MOX). Sixteen rats were contaminated by intramuscular injection of a 1 mg MOX suspension and then treated daily for 7 d with LIHOPO (30 or 200 micromol kg(-1)) or DTPA (30 micromol kg(-1)). LIHOPO was inefficient for removing Pu, Am and U from the wound site. However, it reduced Pu retention in carcass and liver by factors of 2 and 6 respectively, and Am retention in carcass and liver by factors of 10 and 30. In contrast, the effect of LIHOPO on U was to decrease the retention in kidneys by a factor of 75. These results confirm that LIHOPO is a good candidate for use after contamination with MOX, in combination with localised wound lavage or surgical treatment aimed at removing most of the contaminant at the wound site.     

Separation of U and Pu by countercurrent chromatography with support-free liquid stationary phase in the TBP white spirit nitric acid system

The article deals with the use of countercurrent chromatography (CCC), support-free partition chromatography, for separation of U and Pu from the organic extract obtained directly by the dissolution of MOX fuel in supercritical CO₂ containing the complex TBP · nHNO₃. White spirit solutions with various TBP concentrations were used as a stationary phase. The effects of the compositions of the stationary and mobile phases on the U/Pu partition efficiency are studied. The CCC method allows the separation of U and Pu under conditions of a TBP concentration gradient in the stationary phase and also of an HNO₃ concentration gradient in the mobile phase. Chromatographic separation first gives the Pu fraction containing 98.9% of Pu and 0.07% of U, and then the U fraction (99.93% of U and 1.1% of Pu). The separation time is 50 min.     

Separation of Am(III) and Pu(IV) at Their Joint Oxalate Precipitation

A procedure was developed for purification of plutonium to remove americium by oxalate precipitation using diethyl oxalate as a precipitant. The developed procedure was tested under laboratory conditions and on an enlarged installation. It was shown that the decontamination factor of plutonium from americium can exceed 2 ×10³. The plutonium oxalate precipitate can be dissolved by heating with HNO₃ in the presence of V(V) catalyst and also by electrochemical dissolution at alternating current.     

Extraction of Am(III), U(VI), and Pu(IV) from HCl Solutions with Diphenyl(dibutylcarbamoylmethyl)phosphine Oxide

Extraction of Am(III), U(VI), and Pu(IV) from HCl solutions with solutions of diphenyl(dibutylcarbamoylmethyl)phosphine oxide in dichloroethane and m-nitrobenzotrifluoride was studied. The curves of the Am(III) and U(VI) extraction as a function of the acid concentration pass through a minimum at [HCl] = 1 M, followed by a steep ascent at the acid concentration increased further. The logarithmic dependences of the distribution factors on the extractant concentration are nonlinear, with the slope of the upper portions close to 2 for Am(III) and 1 for U(VI). Pu(IV) is extracted noticeably worse than U(VI). A significant anomalous aryl strengthening effect is observed in HCl solutions.     

Complexation of An(VI) (An = U, Np, Pu, Am) with 2,6-pyridinedicarboxylic acid in aqueous solutions. Synthesis and structures of new crystalline compounds of U(VI), Np(VI), and Pu(VI)

Complexation of An(VI) (An = U, Np, Pu, Am) with 2,6-pyridinedicarboxylic (dipicolinic) acid in aqueous solutions was studied. All these actinides form with dipicolinic acid anion, PDC^2? 1: 1 and 1: 2 complexes. The PDC^2? ion coordinates to actinide(VI) ions in solutions in tridentate fashion. In 1: 2 complexes, the f-f transition bands in the electronic absorption spectra are very weak, which is associated with approximate central symmetry of the coordination polyhedron (CP) of the An atom. The apparent stability constants of Pu(VI) complexes were measured in a wide pH range, and the concentration stability constants of An(VI) (An = U, Np, Pu, Am) were determined. The crystalline complexes [Li₂AnO₂(PDC)₂]·2H₂O (An = U, Np, Pu) and [AnO₂(PDC)] _n (An = Np, Pu) were synthesized, and their structures were determined by single crystal X-ray diffraction. The X-ray data confirmed the conclusion that CP of An atoms in the complex ions AnO₂·(PDC) ₂ ^2? is centrosymmetrical. In the isostructural series of [Li₂AnO₂(PDC)₂]·2H₂O, the actinide contraction is manifested in shortening of the An-O distances in the “yl” groups in going from U to Pu.     

Extraction and sorption preconcentration of U(VI), Am(III), and Pu(IV) from nitric acid solutions with alkylenediphosphine dioxides

The extraction of U(VI), Am(III), and Pu(VI) from nitric acid solutions in the form of complexes with alkylenebis(diphenylphosphine) dioxides and their sorption with POLIORGS F-6 sorbent prepared by noncovalent immobilization of methylenebis(diphenylphosphine) dioxide (MDPPD) on a KhAD-7M? polymeric matrix were studied. The preconcentration conditions and distribution coefficients of U(VI), Am(III), and Pu(IV) in their sorption from 3 M HNO₃ were determined. The possibility of concentrating actinides from multicomponent solutions was demonstrated. The composition and nature of complexes of U(VI) with MDPPD were determined from the ³¹P NMR data.     

AnSn Coordination Polyhedra (An = Th,U, Np,Pu, Am,Cm, or Cf) in Crystal Structures

Radiochemistry - Voronoi–Dirichlet polyhedra (VPD) were used to perform a crystal-chemical analysis of 216 sulfides containing in their crystal structures 296 coordination polyhedra AnSn (An...     

Thermodynamic modeling of thermal processes involving actinides (U,Am, Pu) in the course of heating radioactive graphite in steam

The behavior of actinides (U, Am, Pu) in the course of combustion of radioactive graphite in steam was studied by thermodynamic modeling. Thermodynamic modeling was performed using TERRA program in the temperature interval from 373 to 3273 K to determine the possible composition of volatile actinide species formed in the course of graphite utilization by heating in steam. The modeling shows that the actinides at high temperatures are present in the following forms: U, as UO₃ and UO₂ vapors and as UO₃ ^– and UO₂ ⁺ ions; Am, as Am vapor; and Pu, as PuO₂ and PuO vapors and as ionized PuO⁺. The main reactions within separate phases and at the interface were revealed, and their equilibrium constants were determined.     

Extraction of U(VI), Th(IV), Pu(IV), and Am(III) from nitric acid solutions with polyfunctional organophosphorus acids

Extraction of microamounts of U(VI), Th(IV), Pu(IV), and Am(III) nitrates from aqueous HNO₃ solutions with solutions of (diphenylphosphinylmethyl)phenylphosphinic, (di-p-tolylphosphinylmethyl)phenylphosphinic, and (dioctylphosphinylmethyl)phosphinic acids and of butyl hydrogen (diphenylphosphinylmethyl)phosphonate in organic diluents was studied. The metal: extractant stoichiometric ratio in the extractable complexes was determined, and the diluent effect on the extraction efficiency was examined. The possibility of using a macroporous polymeric sorbent impregnated with (dioctylphosphinylmethyl)phenylphosphinic acid for concentrating metal ions from HNO₃ solutions was demonstrated.     

Nondestructive characterisation of MOX fuel rods using gamma autoradiography (GAR)

During the fabrication of mixed uranium-plutonium oxide (MOX) fuel rods, two important characteristics to be checked in as-fabricated fuel pins are plutonium enrichment and plutonium dioxide agglomerates. The mixed oxide fuel pellets are made via mechanical mixing of uranium dioxide and plutonium dioxide powders by cold compaction and sintering. The chance of loading a wrong Pu enrichment pellet and having pellets with plutonium dioxide agglomerates in a fuel pin cannot be ruled out. A simple nondestructive evaluation technique is felt necessary to ensure at the last stage (in the welded pins) to check these two characteristics. During the fabrication of MOX fuel rods for Boiling Water Reactors at Advanced Fuel Fabrication Facility of BARC at Tarapur, Gamma-auto-radiography was successfully used to evaluate composition of MOX pellets and to detect presence of PuO₂ agglomerates in the peripheral region. The fuel pins are allowed to be in contact with industrial X-ray films loaded in cassettes for a long time and the processed films are carefully evaluated. Experiments were made to standardise the conditions for distinguishing fuel pellets of different composition by gamma-auto-radio-graphy of fuel pins loaded with pellets of different composition. Gamma-auto-radiography of fuel pins containing agglomerates of different sizes was also carried out. This paper describes the experimental details of the technique, results obtained and compare with other nondestructive evaluation techniques available.     

Processes for preparing mixed oxides,as applied to conditions of spent nuclear fuel reprocessing without complete separation of U and Pu

Technologies for spent nuclear fuel reprocessing and fabrication of new fuel without complete separation of Pu from U meet the requirements of nonproliferation of nuclear weapons and allow the number of process operations to be reduced. Three types of known processes for converting a joint solution to mixed plutonium and uranium oxides are considered in a brief literature review: “direct“ thermal denitration without isolation of intermediates, thermal denitration with conversion of nitrates to oxides in several steps, and isolation of intermediates with their separation from solution in the first step, followed by their heat treatment. It is advisable to combine processes for preparation of single-phase mixed U-Pu oxides with their purification. Additional experimental data on coprecipitation of uranyl and plutonyl oxalates in oxidizing medium are presented.     

Radiochemical determination of 241Am and Pu(alpha) in environmental materials

Americium-241 and plutonium determinations will become of greater importance over the coming decades as 137Cs and 241Pu decay. The impact of 137Cs on environmental chronology has been great, but its potency is waning as it decays and diffuses. Having 241Am and Pu as unequivocal markers for the 1963 weapon fallout maximum is important for short time scale environmental work, but a fast and reliable procedure is required for their separation. The developed method described here begins by digesting samples using a lithium borate fusion although an aqua regia leachate is also effective in many instances. Isolation of the Am and Pu is then achieved using a combination of extraction chromatography and conventional anion exchange chromatography. The whole procedure has been optimized, validated, and assessed for safety. The straightforwardness of this technique permits the analysis of large numbers of samples and makes 241Am-based techniques for high-resolution sediment accumulation rate studies attractive. In addition, the technique can be employed for the sequential measurement of Pu and Am in environmental surveillance programs, potentially reducing analytical costs and turnround times.     

Sorption-barrier properties of granitoids and andesite-basaltic metavolcanites with respect to Am(III) and Pu(IV): 1. Absorption of Am and Pu from groundwater on monolithic samples of granitoids and andesite-basaltic metavolcanites

A model study is made of the sorption-barrier properties of intact monolithic samples of granitoids and andesite-basaltic metavolcanites with respect to Am(III) and Pu(IV). In sorption from simulated groundwater (pH 8.3), the surface distribution coefficient in surface sorption K _a was determined to be 8–37 and 4–80 cm for Am and Pu, respectively. The mineral components of the rocks responsible for the radionuclide sorption were identified by autoradiography. The rocks tested are characterized by high retention capacity for Am and Pu.     

Dissolution of U,Np, Pu,and Am Oxides and Their Mixtures in Tributyl Phosphate Saturated with HNO3

It was shown for the first time that weighable amounts of uranium dioxide and its mixtures with Np, Pu, and Am in the form of solid solutions can be efficiently and quantitatively dissolved in the presence of tributyl phosphate saturated with HNO₃. Individual PuO₂ and NpO₂ are not dissolved under these conditions. In treatment of a mechanical mixture of UO₂ with PuO₂ or NpO₂, uranium is completely dissolved, while plutonium and neptunium remain in the precipitate.     

Thermodynamic Analysis of the Carbothermal Synthesis of (U,Pu)N

Radiochemistry - Thermodynamic modeling was performed to study and evaluate the behavior of the physicochemical system characteristic of the carbothermal synthesis of uranium and plutonium...     

Sorption recovery of U(VI), Pu(IV), and Am(III) from nitric acid solutions with solid-phase extractants based on Taunit carbon nanotubes and polystyrene supports

Sorption procedures were developed for recovering U(VI), Pu(IV), and Am(III) with solid-phase extractants (SPEs) prepared by impregnation of Taunit carbon nanotubes and polystyrene supports with diphenyl( dibutylcarbamoylmethyl)phosphine oxide (CMPO) and tri-n-octylphosphine oxide (TOPO) The impregnation and actinide recovery were performed in the batch mode and using microcolumns. Procedures for support impregnation and SPE preparation are described. Conditions were found for sorption recovery of U(VI), Pu(IV), and Am(III) from 3 M HNO₃ solutions. The possibility of actinide desorption was demonstrated. The effect of macrocomponents on the degree of actinide recovery was examined.     

Sorption-barrier properties of granitoids and andesite-basaltic metavolcanites with respect to Am(III) and Pu(IV): 2. Absorption of Am and Pu from groundwater on crushed samples of granitoids and andesite-basaltic metavolcanites and also on their rock-forming minerals

A model study is made of the sorption-barrier properties of crushed samples of granitoids and andesite-basaltic metavolcanites with respect to Am(III) and Pu(IV). In sorption from simulated groundwater (pH 8.3), the volume distribution coefficient K _d of Am was determined to be (0.8–1.6) × 10³ and (3.4–7.0) × 10³ cm³ g^?1, and that of Pu, (0.5–1.7) × 10³ and (1.0–10.0) × 10² cm³ g^?1 for metavolcanites and granitoids, respectively, suggesting good sorption-barrier properties of these rocks. The sorption power of the basic rock-forming minerals of granitoids decreases in the order biotite > feldspar > quartz. The results obtained in this study can be used as input data in predicting the rates of Am and Pu migration with groundwater.     

Chemical resolution of Pu+ from U+ and Am+ using a band-pass reaction cell inductively coupled plasma mass spectrometer

Determination of the concentration and distribution of the Pu and Am isotopes is hindered by the isobaric overlaps between the elements themselves and U, generally requiring time-consuming chemical separation of the elements. A method is described in which chemical resolution of the elemental ions is obtained through ion-molecule reactions in a reaction cell of an ICPMS instrument. The reactions of "natural" U(+), (242)Pu(+), and (243)Am(+) with ethylene, carbon dioxide, and nitric oxide are reported. Since the net sensitivities to the isotopes of an element are similar, chemical resolution is inferred when one isobaric element reacts rapidly with a given gas and the isobar (or in this instance surrogate isotope) is unreactive or slowly reactive. Chemical resolution of the m/z 238 isotopes of U and Pu can be obtained using ethylene as a reaction gas, but little improvement in the resolution of the m/z 239 isobars is obtained. However, high efficiency of reaction of U(+) and UH(+) with CO(2), and nonreaction of Pu(+), allows the sub-ppt determination of (239)Pu, (240)Pu, and (242)Pu (single ppt for (238)Pu) in the presence of 7 orders of magnitude excess U matrix without prior chemical separation. Similarly, oxidation of Pu(+) by NO, and nonreaction of Am(+), permit chemical resolution of the isobars of Pu and Am over 2-3 orders of magnitude relative concentration. The method provides the potential for analysis of the actinides with reduced sample matrix separation.     

Determination of 90Sr, 238Pu, 239,240Pu, 241Pu, and 241Am in soils of chernobyl region using various sample preparation techniques

Radiochemical analysis is made of soddy-podzolic sandy and peaty gley soils collected from the Chernobyl zone. Radionuclides were separated by extraction chromatography or ion exchange, and then determined using an α-ray spectrometer and liquid scintillation and proportional gas counters. Leaching from the preliminarily calcined sample with 8 M HNO₃ does not ensure 100% recovery of radionuclides from peaty gley soil.