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The chromatographic resins DGA and LN have been used sequentially to separate curium, californium, plutonium, and various fission products (FPs) to obtain products with a pure radioisotope or a rad...

This work describes a streamlined approach to the recovery and purification of 234U from aged 238Pu. Key modifications to conventional techniques are described that result in an efficient uranium and plutonium separation process that successfully produced tens of grams of ultrahigh purity 234U at Oak Ridge National Laboratory. A new processing scheme was developed using an anion exchange column run in a cycling mode with HNO3 media as the primary uranium purification process. This was followed with a uranyl peroxide precipitation step to purify and solidify the uranium before conversion to U3O8. Both the traditional uranyl peroxide precipitation method and a new cascade uranyl peroxide precipitation process were used for final purification of the 234U.

The separation of Eu and Sm from each other is made difficult by their shared preference to be a trivalent cation as well as their similar ionic radii. A new approach to separating Eu and Sm using ...

The use of Laser Induced Breakdown Spectroscopy as a technique for analysis of europium and samarium concentrations in ceramic aluminum oxide samples was investigated for application to studying lanthanide separation processes. Metal oxide powders of europium and/or samarium were mixed with aluminum oxide at varying concentrations and pressed into pellets. A comparison of different approaches to building a univariate simple linear regression analysis curve is presented. A linear behavior was seen over the total lanthanide concentration range from 0.086 to 12.358 weight percent (wt%). The calculated limits of detection for the univariate calibration curves were determined to range from 0.001 to 0.108 wt% and 0.001 and 0.183 wt% for europium and samarium, respectively.

Purification of berkelium-249 (249Bk) at Oak Ridge National Laboratory requires a multi-step process with Bk Finishing as the final step. The current multi-process takes 6–7 months, including ~ 3 months spent on Bk Finishing. Recent trials of using columns of either LN resin or MP-1 anion exchange resin adsorbing the tetravalent 249Bk4+ in 8M HNO3-0.5M NaBrO3, followed by eluting impurities (e.g. Ce4+, Cf3+) and stripping 249Bk resulted in a new dual column method of selectively separating 249Bk from other trans-curium elements and fission product impurities, which is proposed to replace the current Bk Finishing with higher Bk purification efficiency and shorter processing period. This paper displays the experimental results and describes the dual column method with relevant chemistry issues.

Characterization of samarium, and europium beta”-alumina derivatives has been carried out using x-ray photoelectron spectroscopy (XPS). Beta”-alumina has been widely studied as a material capable of incorporating many different cations into its lattice structure, such as sodium and many of the lanthanide elements. This unique behavior has been recently explored at Oak Ridge National Laboratory for separating samarium and europium from each other. The XPS of samarium and europium in the beta”-alumina structure are reported here. Additionally, the XPS spectra of the europium and samarium trichloride starting materials are presented in the database.

Oxygen exchange reactions performed on PuO 2 suggest the reaction is influenced by at least three mechanisms: an internal chemical reaction, surface mobility of active species/defects, and surface exchange of gaseous oxygen with lattice oxygen. Activation energies for the surface mobility and internal chemical reaction are presented. Determining which mechanism is dominant appears to be a complex function including at least specific surface area and temperature. Thermal exposure may also impact the oxygen exchange reaction by causing reductions in the specific surface area of PuO 2 . Previous CeO 2 surrogate studies exhibit similar behavior, confirming that CeO 2 is a good qualitative surrogate for PuO 2 , in regards to the oxygen exchange reaction. Comparison of results presented here with previous work on the PuO 2 oxygen exchange reaction allows complexities in the previous work to be explained. These explanations allowed new conclusions to be drawn, many of which confirm the conclusions presented here.

A project to determine the neutron-capture cross section of long lived fission products and minor actinides has been started by using a beam-line at Japan Proton Accelerator Research Complex (J-PARC). One of the target nuclides is 93Zr, which was prepared in Oak Ridge National Laboratory. Qualitative and quantitative analyses on the sample were performed at Kyoto University. The isotopic composition of m Zr (m 90, 91, 92, 93, 94, and 96) was precisely determined by multi-collector thermal ionization mass spectrometry with

Greater availability of therapeutic radioisotopes is required to meet the demands for increasing clinical applications in nuclear medicine, oncology and interventional cardiology. Because of the need for very high specific activity products, methods other than direct neutron capture reactions (n,γ-elastic and n,n’-inelastic routes) are required to insure that the highest specific activity - and hopefully no carrier added (nca) - radioisotopes are available. Two major methods to obtain nca radioisotopes from reactors are through the use of radionuclide generator systems using reactor-produced parents and the formation of desired radioisotopes through beta-decay of reactor-produced species. In this paper we describe our recent development of new approaches to obtain nca 177Lu from the decay of reactor-produced 177Yb and free of the long-lived 177mLu (T1/2 = 160 d) radiocontaminant. We also describe preliminary results of our new “indirect” method for the production of 195mPt via decay of 195mIr, reactor-produced by neutron irradiation of highly enriched 193Ir.

Heterogeneous oxidation has been utilized for one-step conversion of metallic W to WO 3 at ∼750 °C in a stream of air, with simultaneous separation of W from Re and Os impurities which are coproduced during the long neutron irradiation of the W target (∼50 days). The freshly produced WO 3 is readily soluble in a base for use in a 188 W/ 188 Re biomedical generator. The metallic tungsten target, in the form of small pellets (4 x 3 mm diameter x height), was prepared by cold pressing W powder at ∼3 tons/cm 2 . Sintering (vacuum annealing) the pellets at 1200-1400 °C, under 1-10 μTorr pressure provided good mechanical strength. Analysis of the black residue, which is typically encountered in the dissolution of neutron-irradiated WO 3 targets, indicated that 186 W and 187 Re were the main constituents. Further, the specific activity of insoluble 188 W in the residue was ∼200 times smaller than that of soluble 188 W.

The pcKa values of acetic acid (HAc) and hydrofluoric acid (HF), as well as pKw values of H2O, were measured in this study at 25 °C and ionic strengths of I = 0.1 ∼ 14 m (NaClO4). Potentiometric titration was the experimental technique for these measurements using either a glass pH electrode or a fluoride ion selective electrode.