Jonathan D. Burns

Separation of Americium from Curium by Oxidation and Ion Exchange

Nuclear energy has the potential to be a clean alternative to fossil fuels, but in order for it to play a major role in the US, many questions about the back end of the fuel cycle must be addressed. One of these questions is the difficult separation of americium from curium. Here, we report the oxidation of Am in two systems, perchloric acid and nitric acid and the affect of changing the acid has on the oxidation. K(d) values were observed and a direct separation factor was calculated and was seen to be as high as 20 for four metal(IV) pillared phosphate phosphonate inorganic organic hybrid ion exchange materials. These ion exchangers are characterized by very low selectivity for cations with low charge but extremely high uptake of ions of high charge.

Rates of Exchange of Cs+ and Sr2+ for Poorly Crystalline Sodium Titanium Silicate (CST) in Nuclear Waste Systems

The compound sodium titanium silicate, popularly known as CST, is highly selective for Cs+. It was synthesized for the purpose of removing 137Cs from basic nuclear waste systems. This compound has a tunnel structure in which the Cs+ ion just fits but diffusion through the tunnels is relatively slow. CST loses its ability to sequester Cs+ in the strongly basic nuclear waste solutions. However, replacement of titanium with 25 mol% of niobium increases the selectivity to a satisfactory level. It has been found that producing a less crystalline form of Nb-CST greatly improves the rate of Cs+ removal. Additionally, the non-niobium CST is selective for strontium both as Sr2+ and Sr(OH)+. It is suggested that both radioisotopes of cesium and strontium may be efficiently extracted by a combination of a mixture of poorly crystalline CST and Nb-CST.

Group Hexavalent Actinide Separations: A New Approach to Used Nuclear Fuel Recycling.

Hexavalent Np, Pu, and Am individually, and as a group, have all been cocrystallized with UO2(NO3)2·6H2O, constituting the first demonstration of an An(VI) group cocrystallization. The hexavalent dioxo cations of Np, Pu, and Am cocrystallize with UO2(NO3)2·6H2O in near proportion with a simple reduction in temperature, while the lower valence states, An(III) and An(IV), are only slightly removed from solution. A separation of An(VI) species from An(III) ions by crystallization has been demonstrated, with an observed separation factor of 14. Separation of An(VI) species from key fission products, (95)Zr, (95)Nb, (137)Cs, and (144)Ce, has also been demonstrated by crystallization, with separation factors ranging from 6.5 to 71 in the absence of Am(VI), while in the presence of Am(VI), the separation factors were reduced to 0.99-7.7. One interesting observation is that Am(VI) shows increased stability in the cocrystallized form, with no reduction observed after 13 days, as opposed to in solution, in which >50% is reduced after only 10 days. The ability to cocrystallize and stabilize hexavalent actinides from solution, especially Am(VI), introduces a new separations approach that can be applied to closing the nuclear fuel cycle.

Samarium electrodeposited acetate and oxide thin films on stainless steel substrate characterized by XPS

Characterization of a samarium thin film deposited on a stainless steel substrate using electrodeposition was carried out with a Thermo Scientific K-Alpha X-ray photoelectron spectrometer. Two types of samarium electrodeposition samples were studied, one as-deposited and one heated to 700 °C in an air atmosphere. Survey scans include peaks coming from the stainless steel substrate, such as Fe and Cr. An x-ray photoelectron spectroscopy (XPS) survey spectrum, Sm 3d, C 1s, and O 1s narrow scans are shown. It was determined, using XPS, that the heating process decomposed the deposited Sm acetate to Sm2O3.

Californium Recovery from Palladium Wire

The recovery of 252Cf from palladium-252Cf cermet wires was investigated to determine the feasibility of implementation into the cermet wire production operation at the Radiochemical Engineering Development Center at Oak Ridge National Laboratory. The dissolution of Pd wire in 8 M HNO3 and trace amounts of HCl was studied at both ambient and elevated temperatures. These studies showed that it took days to dissolve the wire at ambient temperature and only 2 hours at 60°C. Adjusting the ratio of the volume of solvent to the mass of the wire segment from 0.176 mL/mg down to 0.019 mL/mg resulted in little change in the kinetics of dissolution. A successful separation of 153Gd, a surrogate for 252Cf, from Pd was demonstrated using ion exchange chromatography.

Nuclear Forensics Methodology for Reactor-Type Attribution of Chemically Separated Plutonium

Abstract A nuclear forensics methodology has been developed that is capable of source attribution of separated weapons-grade plutonium in case of an interdiction. The methodology utilizes plutonium and contaminant fission product isotopes within the separated plutonium sample to determine the characteristics (reactor parameters) of the interdicted material. The reactor parameters of interest include source reactor type, fuel irradiation burnup, and time since irradiation. The MCNPX-2.7 radiation transport code was used to model reactor cores and perform neutronics simulations to estimate the resulting isotopes of irradiated UO2 fuel. The simulation results were used to create a reactor-dependent library of irradiated fuel isotope ratio values as a function of fuel burnup and time since irradiation. Ratios of intra-element isotopes (fission product or actinide) are used as characteristics to determine a combination of reactor parameters of interest that could have produced the interdicted sample. The isotopes selected for the attribution methodology development were based upon the initial criteria of isotope production yield in fuel and half-life. Subsequently, intra-element isotope ratios were formed with the criterion that the ratio must have a functional dependence on at least one of the reactor parameters of interest. The developed methodology compares the values of reactor-dependent intra-element isotope ratios in the library developed to the same ratios of the interdicted sample. A maximum likelihood calculation methodology was utilized to perform the aforementioned multiple intra-element isotope ratio comparison to produce a single metric to depict the result of the comparison. The methodology can predict the reactor type, fuel burnup, and time since irradiation of the sample by selecting the array of reactor-dependent intra-element isotope ratios that provides the maximum likelihood value. The methodology was tested with intra-element ratios of pseudo interdicted sample data and found to be viable for source attribution.

Sensitivity Studies and Experimental Evaluation for Optimizing Transcurium Isotope Production

Abstract This work applies to recent initiatives at the Radiochemical Engineering Development Center at Oak Ridge National Laboratory to optimize the production of transcurium isotopes in the High Flux Isotope Reactor in such a way as to prolong the use of high-quality heavy curium feedstock. By studying the sensitivity of fission and transmutation reaction rates to the neutron flux energy spectrum, a flux filtering methodology is explored for increasing the fraction of (n,γ) reactions per neutron absorption. Filter materials that preferentially absorb neutrons at energies considered detrimental to optimal transcurium production are identified, and transmutation rates are examined with high-energy resolution. Experimental capsules are irradiated employing filter materials, and the resulting fission and activation products are studied to validate the filtering methodology. Improvement is seen in the production efficiency of heavier curium isotopes in 244Cm and 245Cm targets and potentially in the production of 252Cf from mixed californium targets. Further analysis is recommended to evaluate longer-duration irradiations more representative of typical transcurium production.