Volume 102, Issue 12

The occurrence of plutonium dioxide (PuO 2 ) either from direct deposition or from the precipitation of plutonium-bearing solutions in contaminated soils and sediments is well described, particularly for the Hanford site in Washington State. However, past research has suggested that plutonium at the Hanford site may exist in chemical forms in addition to PuO 2 . Although the majority of the plutonium is present as oxide, we present evidence for the formation of nano-sized mixed plutonium- iron phosphate hydroxide structurally related to the rhabdophane group minerals in 216-Z9 crib sediments from Hanford using both transmission electron microscopy (TEM) and X-ray absorption spectroscopy (XAS). The iron-plutonium phosphate formation may depend on the local microenvironment in the sediments, availability of phosphate, and hence the distribution of these minerals may control long-term migration of plutonium in the soil.

Electrochemistry of NpO 2 2+ /NpO 2 + couple in 1 M H 2 SO 4 was studied on bare and modified glassy carbon (GC) electrodes by cyclic voltammetry (CV), differential pulse voltammetry (DPV) and electrochemical impedance spectroscopy (EIS). The modified electrode (SWCNTs-GC) was characterized by scanning electron microscopy (SEM). Electrocatalysis of NpO 2 2+ /NpO 2 + redox reaction was observed on SWCNT-GC electrode. The lower charge transfer resistance at SWCNT-GC reflects faster rate of electron transfer reaction of Np(VI)/Np(V) couple. These results are interesting to develop electroanalytical methodologies for neptunium determination using SWCNT modified electrode. To the best of our knowledge, this is the first study on the electrocatalysis of neptunium on SWCNT modified electrode.

To develop an aqueous reprocessing process for spent fuel of fast reactor, a counter current experiment (U-Pu co-stripping) based on a simplified PUREX process was carried out with 30%TBP/Kerosene solution containing U, Pu and HNO 3 . The experiment was done in a 16-stage miniature centrifugal contactor. The stripping percentage of U was 99.99%, Pu 99.94%. U stripping mainly took place in X1-feeding stage (11.00 mol/L HNO 3 feeding stage) to the 16th one, while Pu stripping mainly took place in the first stage to the X1-feeding stage. And a simple method was presented to calculate c U/X1−feeding stage, Org when the X1-feeding stage number is more than 7 and the stripping percentage of U from the X1-feeding stage is more than 99.9% through the rear stages' stripping. The calculation results by this method were validated by Computer Simulation when only changing U concentrations in feed solution from 40.0 g/L to 90.0 g/L or only changing the X1-feeding stage from the 7th stage to the 11th one. Finally the optimized technology was given.

Qualitative and quantitative understanding of Po volatilization under different conditions is of key importance for safety assessments of lead-bismuth eutectic (LBE) based nuclear reactors, spallation targets and accelerator driven systems. In this work we explore the possibilities of the transpiration method in combination with simple models to study the equilibrium and kinetics of Po evaporation from highly diluted solutions in lead-bismuth eutectic between 600 and 1000 ℃ in Ar/5% H 2 and Ar. On the basis of evaporation experiments at various carrier gas flow rates, we identified the conditions of vapor saturation allowing the determination of equilibrium constants. From the limiting behavior at high flow rates, values for the maximal evaporation rate of Po from LBE were estimated. Measurements of evaporation as a function of time were consistent with the assumption that polonium dissolved in LBE obeys Henry's law. A theoretical analysis furthermore suggested that diffusion of polonium in LBE was not a rate limiting factor for evaporation under vapor saturation conditions. Newly determined values for the Henry constant of Po in LBE between 600 and 1000 ℃ were consistent with previously derived correlations.

We report on the in-situ synthesis of metal carbonyl complexes with short-lived isotopes of transition metals. Complexes of molybdenum, technetium, ruthenium and rhodium were synthesized by thermalisation of products of neutron-induced fission of 249 Cf in a carbon monoxide-nitrogen mixture. Complexes of tungsten, rhenium, osmium, and iridium were synthesized by thermalizing short-lived isotopes produced in 24 Mg-induced fusion evaporation reactions in a carbon monoxide containing atmosphere. The chemical reactions took place at ambient temperature and pressure conditions. The complexes were rapidly transported in a gas stream to collection setups or gas phase chromatography devices. The physisorption of the complexes on Au and SiO 2 surfaces was studied. We also studied the stability of some of the complexes, showing that these start to decompose at temperatures above 300 ℃ in contact with a quartz surface. Our studies lay a basis for the investigation of such complexes with transactinides.

In order to update the recommended cross section data of IAEA for production of 111 In and 114 m In through the 112 Cd(p,2n) and 114 Cd(p,n) reactions new measurements were performed. In stacked-foil irradiations with incident proton energy of 36.7 and 25 MeV on highly enriched 112 Cd and 114 Cd targets, the excitation functions for 109,110 g ,110 m ,111,113 m ,114 m In were determined, relative to the monitor reactions nat Cu(p,x) 62,65 Zn. The results are compared with the available literature values (also extracted from measurements on nat Cd) and with data listed in the on-line library TENDL-2012 (calculated with the TALYS 1.4 theoretical code). The industrial, PC-controlled automated dissolution and separation chemistry apparatus for delivery of large quantities of nca 111 In and the recovery of enriched 112 Cd from processed irradiated targets are described.

A DTPA-conjugated amlodipine analog (DTPA-AMLO) 3 , was prepared for possible voltage gated calcium channel imaging after radiolabeling with Ga-67. [ 67 Ga]-DTPA-AMLO complex was prepared starting [ 67 Ga]gallium chloride and DTPA-AMLO in 60–90 min at 50–60 ℃ in phosphate buffer. The partition co-efficient and stability of the tracer was determined in final solution (25 ℃) and presence of human serum (37 ℃) up to 24 h. The biodistribution of the labeled compound in wild-type rats were determined up to 72 h using organ counting and SPECT. The radiolabled complex was prepard in high radiochemical purity (>96%, RTLC and >98% HPLC) and significant specific activity (7–10 GBq/mmol). The log P for the complex was calculated as − 0.594, consistent with a water soluble complex. The tracer is mostly washed out through kidneys which were in full compliance with the amlodipine metabolism and imaging studies demonstrated the same behavior. The tracer uptake in organs with smooth muscles was observed in stomach, colon as well as intestine.

In this study, an alkaline-digestion pretreatment and a subsequent ICP-MS measurement were conducted for iodine-129 (I-129) determination in low-level radioactive waste. A TMAH + H 2 O 2  + Triton X-100 mixed alkaline digestion was the most effective mixture for I-129 determination. Using this alkaline reagent, a high level of I-129 recovery (101 ± 6%) was achieved for the analysis of the I-129-spiked standard reference materials NIST 2709 and 2711. Importantly, the I-129 concentrations determined for ten real samples provided by the Lan-Yu radioactive waste temporary storage site were found to be below the detection limit (0.011 mg/kg). This value was only approximately 30–70% of the values determined using the I-129/Cs-137 scaling factor. This means that using the I-129/Cs-137 scaling factor severely overestimates the I-129 concentration in these low-level radioactive wastes. We therefore suggest that a detailed re-inspection of the I-129/Cs-137 scaling factor should be performed to appropriately categorize these low-level radioactive wastes.