Volume 101, Issue 5 -

We report on our attempt to reproduce the 228 Th decay enhancement under the effect of power ultrasound claimed recently in the literature. The evolution of natural thorium isotopic composition ( 232 Th, 230 Th, and 228 Th) in various aqueous solutions submitted to 20 kHz power ultrasound (0.18 W mL −1 , 20 ºC, Ar). The radioactivities of the three isotopes of thorium ( 232 Th, 230 Th and 228 Th) were followed by α liquid scintillation counting technique. No change was detected during ultrasonic irradiation. Our results demonstrate that under the studied conditions the rate of Th isotopes radioactive decay is not modified under the effect of ultrasound on aqueous solutions of thorium salts.

The complexation of Cm(III) and Eu(III) with 2,6-bis(5-(2,2-dimethylpropyl)-1H-pyrazol-3-yl)pyridine (C5-BPP) and 2-bromohexanoic acid as lipophilic anion has been investigated by time-resolved laser fluorescence spectroscopy. Upon increasing ligand concentration three different Cm(III)-C5-BPP species with emission bands at 604.1 nm, 607.9 nm and 611.4 nm, respectively, are found and attributed to complexes with one, two and three C5-BPP molecules in the inner coordination sphere. Comparison with results of TRLFS experiments without 2-bromohexanoic acid shows that the C5-BPP ligand is able to completely displace the lipophilic anion from the inner coordination sphere, forming [Cm(C5-BPP) 3 ] 3+ complexes. This complex is also found in the organic phase of an extraction experiment performed with Cm(III), demonstrating that the lipophilic anion required for the extraction is not directly coordinated to the metal ion in the species formed during extraction. In case of Eu(III) the number of different species formed cannot be determined accurately. Nevertheless, the formation of the complex [Eu(C5-BPP) 3 ] 3+ in the presence of 2-bromohexanoic acid is confirmed.

This paper concerns quantitative isotopic analysis of Am, Cm and Nd contained in an irradiated AmO 1.62 /MgO pellet. The complete analysis protocol is described, from dissolution of the pellets in a shielded line to the laboratory glove separation processes box for TIMS analysis. Emphasis is placed on the separation processes: by ion exchange resin in a hot cell and by HPLC in the laboratory. Intermediate measurements by X-ray fluorescence, alpha spectrometry, and ICP-AES are described.

A solution of 0.1 M N , N -di-2-ethylhexyl- N ´, N ´-di-octyl-3-oxa-1,5-diamide (DEHDODGA)-0.5 M N , N -dihexyloctanamide (DHOA) in n -dodecane has been evaluated for the separation of trivalent actinides from high-level liquid waste (HLLW). The extraction and stripping behaviour of Am(III) and other metal ions present in HLLW was studied. The distribution ratio of various metal ions was measured as a function of concentration of nitric acid and interfering ion. The extraction of metal ions decreased in the order M 4+ ≥ M 3+ ≫ M 2+ ∼ MO 2 2+ ≫ M + . The extraction of Am(III), Eu(III), Y(III), and Zr(IV) was quantitative. However, the distribution ratio of unwanted metal ions such as Fe(III), Co(II), Sb(III), Mn(II), and Cs(I) was negligible. The distribution ratios of Cd(II), Ru(III), Pd(II), Mo(VI), Cr(VI), Ba(II), Ni(II), and Sr(II) were not insignificant, but were quite low. The conditions needed for the quantitative recovery of trivalents from the loaded organic phase were optimized with the use of diethylenetriaminepentaacetic acid (DTPA)-citric acid at pH 3 to facilitate the mutual separation of lanthanides-actinides in the subsequent step.

In the present work the first direct measurement of hydrolysis products of Pu(VI) was achieved by means of nano-electrospray ionization time-of-flight mass-spectrometry. The results indicate that monomeric PuO 2 (OH) + and dimeric (PuO 2 ) 2 (OH) 2 2+ species are present in solution. A trimeric species does not appear within the detection limit of the experiment, in contrast to U(VI) hydrolysis. The relative abundances of the Pu(VI) hydrolysis species in the ESI mass spectra are in good agreement with the published formation constants.

Between 1944 and 1989, 54.5 metric tons of the United States´ weapons-grade plutonium and an additional 12.9 metric tons of fuels-grade plutonium were produced in and separated from irradiated uranium metal fuel at the Hanford Site. Acidic high-activity wastes containing around 600 kg of plutonium were made alkaline and discharged to underground storage tanks from separations, isolation, and recycle processes to yield average plutonium concentration of about 0.003 g/L (or ∼ 0.0002 wt. %) in the ∼ 200 million liter tank waste volume. The plutonium is largely associated with low-solubility metal hydroxide/oxide sludges where its low concentration and intimate mixture with neutron-absorbing elements ( e.g. , iron) are credited in nuclear criticality safety. However, concerns have been expressed that plutonium, in the form of hydrated plutonium oxide, PuO 2 · x H 2 O, could undergo sufficient crystal growth through dissolution and reprecipitation in the alkaline tank waste to potentially become separable from neutron absorbing constituents by settling or sedimentation. Thermodynamic considerations and laboratory studies of systems chemically analogous to tank waste show that the plutonium, precipitated in the alkaline tank waste by neutralization from acid solution, probably entered as 2–5-nm PuO 2 · x H 2 O crystallite particles that, because of the low concentration of the neutral Pu(IV) dissolved species and opposition from radiolytic processes, grow from that point at exceedingly slow rates.

Redox behaviour of Tc(VII)/Tc(IV) was investigated in 0.1 M NaCl solutions containing different reducing agents in the pH range 2 to 13 at 22 ºC under inert Ar atmosphere. In several samples, the 1 × 10 5  mol/dm 3 (M) initially added TcO 4 - was reduced to form a Tc(IV) oxide solid phase with low solubility. The observed Tc redox transformation processes are systematized according to E h -pH conditions in solution, indicating that a borderline for the reduction of Tc(VII) to Tc(IV), TcO 4 - + 3e - + 4H + ⇔TcO 2 · x H 2 O(coll, hyd) + (2-x)H 2 O exists, independent of the reducing chemical system. This experimentally derived borderline is about 100 mV lower than the equilibrium line calculated from the reported standard redox potential of TcO 2 · 1.6H 2 O(s). This behaviour can be related to the existence of more soluble solid phase modifications, i.e. nanoparticulate Tc(IV) oxide species (TcO 2 · x H 2 O(coll, hyd)). The reaction kinetics likewise correlate to the redox potential measured in solution. Slow reduction of Tc(VII) to Tc(IV) was observed when the redox potential in the system was slightly below the above mentioned reduction borderline. Fast reduction was observed in the systems far below the borderline, but also in those systems containing Fe(II) solids, suggesting a specific surface mediated effect in the reduction process. EXAFS analysis on two magnetite samples indicate reduced Tc(IV) species which do not remain adsorbed at the reactive mineral surface and are incorporated in the magnetite structure.

44 Sc is a promising β + -emitter for molecular imaging with intermediate half-life of 4 h. Due to the chemical similarity of Sc 3+ to the Lu 3+ and Y 3+ cations, 44 Sc-DOTA bioconjugates are expected to demonstrate similar properties in vivo as the 177 Lu- and 90 Y-bioconjugates, what is important in planning the radionuclide therapy. 44 Sc can be obtained from the 44 Ti / 44 Sc generator. An alternative method for 44 Sc production can be the irradiation of 44 Ca target at small cyclotrons. The aim of our work was to optimize the parameters of 44 CaCO 3 irradiation and to develop a simple procedure for 44 Sc separation from the calcium target. For optimization study, 44 CaCO 3 targets were irradiated by protons in the energy range of 5.6–17.5 MeV with 9 MeV being found to be the best energy for 44 Ca irradiations. A simple and fast separation procedure of 44 Sc from calcium target was developed using chelating resin Chelex 100. DOTATATE conjugate was successfully radiolabelled with high yield at elevated temperature using the produced 44 Sc. While 44 CaCO 3 is relatively expensive, the cost of 44 Sc-DOTATATE production can be reduced by target recovery. Due to low proton energy required to produce GBq activity level of 44 Sc, the availability of 44 Sc radioisotope could be enhanced to open new opportunities for applications in medical imaging.

Rhenium-186g ( T 1/2 = 89.2 h) is a  β − emitter suitable for therapeutic applications. Current production methods rely on reactor production via 185 Re( n , γ ) which results in low specific activities, thereby limiting its use. Production by p , d activation of enriched 186 W results in a 186g Re product with a higher specific activity, allowing it to be used for targeted therapy with limited receptors. A test target consisting of pressed, sintered nat WO 3 was proton irradiated at Los Alamos (LANL-IPF) to evaluate product yield and impurities, irradiation parameters and wet chemical Re recovery for proof-of-concept for bulk production of 186g Re. We demonstrated isolation of 186g Re in 97% yield from irradiated nat WO 3 targets within 12 h of end of bombardment (EOB) via an alkaline dissolution followed by anion exchange. The recovery process has potential for automation, and WO 3 can be easily recycled for recurrent irradiations. A  186g Re batch yield of 42.7 ± 2.2 μCi/μAh or 439 ± 23 MBq/C was obtained after 24 h in an 18.5 μA proton beam. The target entrance energy was determined to be 15.6 MeV. The specific activity of 186g Re at EOB was measured to be 1.9 kCi (70.3 TBq) mmol −1 , which agrees well with the result of a previous 185,186m Re co-production EMPIRE and TALYS modeling study assuming similar conditions. Utilizing enriched 186 WO 3 , we anticipate that a proton beam of 250 μA for 24 h will provide batch yields of 256 mCi (9.5 GBq) of 186g Re at EOB with specific activities even higher than 1.9 kCi (70.3 TBq) mmol −1 , suitable for therapy applications.